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Sample records for safety manual volume

  1. Highway Safety Program Manual: Volume 3: Motorcycle Safety.

    Science.gov (United States)

    National Highway Traffic Safety Administration (DOT), Washington, DC.

    Volume 3 of the 19-volume Highway Safety Program Manual (which provides guidance to State and local governments on preferred highway safety practices) concentrates on aspects of motorcycle safety. The purpose and specific objectives of a State motorcycle safety program are outlined. Federal authority in the highway safety area and general policies…

  2. Highway Safety Program Manual: Volume 8: Alcohol in Relation to Highway Safety.

    Science.gov (United States)

    National Highway Traffic Safety Administration (DOT), Washington, DC.

    Volume 8 of the 19-volume Highway Safety Program Manual (which provides guidance to State and local governments on preferred highway safety practices) concentrates on alcohol in relation to highway safety. The purpose and objectives of the alcohol program are outlined. Federal authority in the area of highway safety and general policies regarding…

  3. Health and safety manual

    International Nuclear Information System (INIS)

    1980-02-01

    The manual consists of the following chapters: general policies and administration; the Environmental Health and Safety Department; the Medical Services Department: biological hazards; chemical safety; confined space entry; cryogenic safety; electrical safety; emergency plans; engineering and construction; evacuations, trenching, and shoring; fire safety; gases, flammable and compressed; guarding, mechanical; ladders and scaffolds, work surfaces; laser safety; materials handling and storage; noise; personal protective equipment; pressure safety; radiation safety, ionizing and non-ionizing; sanitation; seismic safety; training, environmental health and safety; tools, power and hand-operated; traffic and transportation; and warning signs and devices

  4. Idaho Safety Manual.

    Science.gov (United States)

    Idaho State Dept. of Education, Boise. Div. of Vocational Education.

    This manual is intended to help teachers, administrators, and local school boards develop and institute effective safety education as a part of all vocational instruction in the public schools of Idaho. This guide is organized in 13 sections that cover the following topics: introduction to safety education, legislation, levels of responsibility,…

  5. Missouri Highway Safety Manual Recalibration

    Science.gov (United States)

    2018-05-01

    The Highway Safety Manual (HSM) is a national manual for analyzing the highway safety of various facilities, including rural roads, urban arterials, freeways, and intersections. The HSM was first published in 2010, and a 2014 supplement addressed fre...

  6. Sodium safety manual

    International Nuclear Information System (INIS)

    Hayes, D.J.; Gardiner, R.L.

    1980-09-01

    The sodium safety manual is based upon more than a decade of experience with liquid sodium at Berkeley Nuclear Laboratories (BNL). It draws particularly from the expertise and experience developed in the course of research work into sodium fires and sodium water reactions. It draws also on information obtained from the UKAEA and other sodium users. Many of the broad principles will apply to other Establishments but much of the detail is specific to BNL and as a consequence its application at other sites may well be limited. Accidents with sodium are at best unpleasant and at worst lethal in an extremely painful way. The object of this manual is to help prevent sodium accidents. It is not intended to give detailed advice on specific precautions for particular situations, but rather to set out the overall strategy which will ensure that sodium activities will be pursued safely. More detail is generally conveyed to staff by the use of local instructions known as Sodium Working Procedures (SWP's) which are not reproduced in this manual although a list of current SWP's is included. Much attention is properly given to the safe design and operation of larger facilities; nevertheless evidence suggests that sodium accidents most frequently occur in small-scale work particularly in operations associated with sodium cleaning and special care is needed in all such cases. (U.K.)

  7. Eddy current manual, volume 2

    International Nuclear Information System (INIS)

    Cecco, V.S.; Van Drunen, G.; Sharp, F.L.

    1984-09-01

    This report on eddy current testing is divided into three sections: (a) Demonstration of Basic Principles, (b) Practical (Laboratory) Tests and, (c) Typical Certification Questions. It is intended to be used as a supplement to ΣEddy Current Manual, Volume 1Σ (AECL-7523) during CSNDT Foundation Level II and III courses

  8. Chemistry laboratory safety manual available

    Science.gov (United States)

    Elsbrock, R. G.

    1968-01-01

    Chemistry laboratory safety manual outlines safe practices for handling hazardous chemicals and chemistry laboratory equipment. Included are discussions of chemical hazards relating to fire, health, explosion, safety equipment and procedures for certain laboratory techniques and manipulations involving glassware, vacuum equipment, acids, bases, and volatile solvents.

  9. MARS CODE MANUAL VOLUME III - Programmer's Manual

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Hwang, Moon Kyu; Jeong, Jae Jun; Kim, Kyung Doo; Bae, Sung Won; Lee, Young Jin; Lee, Won Jae

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This programmer's manual provides a complete list of overall information of code structure and input/output function of MARS. In addition, brief descriptions for each subroutine and major variables used in MARS are also included in this report, so that this report would be very useful for the code maintenance. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  10. Tank farms criticality safety manual

    International Nuclear Information System (INIS)

    FORT, L.A.

    2003-01-01

    This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR-), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR- 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type

  11. Environment, safety, and health manual, closeout report

    International Nuclear Information System (INIS)

    1975-12-01

    A coordination draft of the Environment, Safety, and Health (ES and H) manual was submitted on 2 September 1975. Comments provided by Operational Safety personnel were being incorporated by a task team when the effort was terminated on 31 October 1975. This report documents the development history of the manual and provides a status of the manual up to the time the efforts were discontinued. Also discussed are issues which effect completion of the manual. Additionally a plan for completion of the manual is suggested

  12. Quality assurance manual: Volume 1

    International Nuclear Information System (INIS)

    Oijala, J.E.

    1988-06-01

    Stanford Linear Accelerator Center (SLAC) is a DOE-supported research facility that carries out experimental and theoretical research in high energy physics and developmental work in new techniques for particle acceleration and experimental instrumentation. The purpose of this manual is to describe SLAC quality assurance policies and practices in various parts of the Laboratory

  13. Occupational Safety and Health Curriculum Manual.

    Science.gov (United States)

    Gourley, Frank A., Jr., Comp.

    With the enactment of the Occupational Safety and Health Act of 1970, the need for manpower development in the field of industrial safety and hygiene has resulted in the development of a broad based program in Occupational Safety and Health. The manual provides information to administrators and instructors on a program of study in this field for…

  14. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

    On June 27, 1989, the Secretary of Energy announced a 1O-Point Initiative to strengthen environment,safety, and health (ES ampersand H) programs, and waste management activities at involved conducting DOE production, research, and testing facilities. One of the points independent Tiger Team Assessments of DOE operating facilities. The Office of Special Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This volume contains appendices to the Environment, Safety and Health Progress Assessment Manual

  15. Data collection system. Volume 1, Overview and operators manual; Volume 2, Maintenance manual; Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Caudell, R.B.; Bauder, M.E.; Boyer, W.B.; French, R.E.; Isidoro, R.J.; Kaestner, P.C.; Perkins, W.G.

    1993-09-01

    Sandia National Laboratories (SNL) Instrumentation Development Department was tasked by the Defense Nuclear Agency (DNA) to record data on Tektronix RTD720 Digitizers on the HUNTERS TROPHY field test conducted at the Nevada Test Site (NTS) on September 18, 1992. This report contains a overview and description of the computer hardware and software that was used to acquire, reduce, and display the data. The document is divided into two volumes: an overview and operators manual (Volume 1) and a maintenance manual (Volume 2).

  16. Safety Training: "Manual Handling" course in September

    CERN Multimedia

    Safety Training, HSE Unit

    2016-01-01

    The next "Manual Handling" course will be given, in French, on 26 September 2016. This course is designed for anyone required to carry out manual handling of loads in the course of their work.   The main objective of this course is to adopt and apply the basic principles of physical safety and economy of effort. There are places available. If you are interested in following this course, please fill an EDH training request via our catalogue. 

  17. INTRA/Mod3.2. Manual and code description. Volume 2 - User's manual

    International Nuclear Information System (INIS)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the physical modelling of INTRA and the ruling numerics, and volume II, the User's Manual is an input description. This document, the User's Manual, Volume II, contains a detailed description of how to use INTRA, how to set up an input file, how to run INTRA and also post-processing

  18. INTRA/Mod3.2. Manual and code description. Volume 2 - User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the physical modelling of INTRA and the ruling numerics, and volume II, the User`s Manual is an input description. This document, the User`s Manual, Volume II, contains a detailed description of how to use INTRA, how to set up an input file, how to run INTRA and also post-processing

  19. Radiological Emergency Response Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

    D. R. Bowman

    2001-05-01

    This manual was created to provide health and safety (H&S) guidance for emergency response operations. The manual is organized in sections that define each aspect of H and S Management for emergency responses. The sections are as follows: Responsibilities; Health Physics; Industrial Hygiene; Safety; Environmental Compliance; Medical; and Record Maintenance. Each section gives guidance on the types of training expected for managers and responders, safety processes and procedures to be followed when performing work, and what is expected of managers and participants. Also included are generic forms that will be used to facilitate or document activities during an emergency response. These ensure consistency in creating useful real-time and archival records and help to prevent the loss or omission of information.

  20. IAEA publishes first health and safety manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-01-15

    A 'Manual on the Safe Handling of Radioisotopes' was published in English on 15 Dec ember 1958 by the International Atomic Energy Agency. This is a comprehensive handbook of internationally compiled recommendations for users of radioisotopes. It covers organizational, medical and technical aspects of radiation safety practices. It is also the Agency's first technical publication. French, Russian and Spanish editions will appear shortly. The Manual should prove useful to all users of radioisotopes in industry, medicine, research, etc., but is directed mainly to small scale users who may not have access to other sources of information. The recommendations apply only to radioactivity surpassing the limit of 0,002 microcurie concentration per gram of material; or a total activity of more than 0,1 microcuries in the working areas; this limit is based on the most dangerous radioisotopes. The experts state that the limiting level might be higher for less dangerous isotopes, but recommend that all be treated as potentially dangerous. This would have educational value and avoid accidents caused by misidentification. The Manual also stressed that good radiation safety practices depend on effective organization and warns that even very competent workers sometimes ignore or forget important health and safety requirements.

  1. IAEA publishes first health and safety manual

    International Nuclear Information System (INIS)

    1959-01-01

    A 'Manual on the Safe Handling of Radioisotopes' was published in English on 15 Dec ember 1958 by the International Atomic Energy Agency. This is a comprehensive handbook of internationally compiled recommendations for users of radioisotopes. It covers organizational, medical and technical aspects of radiation safety practices. It is also the Agency's first technical publication. French, Russian and Spanish editions will appear shortly. The Manual should prove useful to all users of radioisotopes in industry, medicine, research, etc., but is directed mainly to small scale users who may not have access to other sources of information. The recommendations apply only to radioactivity surpassing the limit of 0,002 microcurie concentration per gram of material; or a total activity of more than 0,1 microcuries in the working areas; this limit is based on the most dangerous radioisotopes. The experts state that the limiting level might be higher for less dangerous isotopes, but recommend that all be treated as potentially dangerous. This would have educational value and avoid accidents caused by misidentification. The Manual also stressed that good radiation safety practices depend on effective organization and warns that even very competent workers sometimes ignore or forget important health and safety requirements

  2. DOE explosives safety manual. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    This manual prescribes the Department of Energy (DOE) safety rules used to implement the DOE safety policy for operations involving explosives. This manual is applicable to all DOE facilities engaged in operations of development, manufacturing, handling, storage, transportation, processing, or testing of explosives, pyrotechnics and propellants, or assemblies containing these materials. The standards of this manual deal with the operations involving explosives, pyrotechnics and propellants, and the safe management of such operations. The design of all new explosives facilities shall conform to the requirements established in this manual and implemented in DOE 6430.1A, ``General Design Criteria Manual.`` It is not intended that existing physical facilities be changed arbitrarily to comply with these provisions, except as required by law. Existing facilities that do not comply with these standards may continue to be used for the balance of their functional life, as long as the current operation presents no significantly greater risk than that assumed when the facility was originally designed and it can be demonstrated clearly that a modification to bring the facility into compliance is not feasible. However, in the case of a major renovation, the facility must be brought into compliance with current standards. The standards are presented as either mandatory or advisory. Mandatory standards, denoted by the words ``shall,`` ``must,`` or ``will,`` are requirements that must be followed unless written authority for deviation is granted as an exemption by the DOE. Advisory standards denoted by ``should`` or ``may`` are standards that may be deviated from with a waiver granted by facility management.

  3. Development of an environmental safety case guidance manual

    International Nuclear Information System (INIS)

    Wellstead, Matthew John

    2014-01-01

    NDA RWMD is currently considering the scope, purpose and structure of a safety case manual that covers the development of nuclear operational, transport and environmental safety cases for a geological disposal facility in the United Kingdom. This paper considers the Environmental Safety Case (ESC) input into such a manual (herein referred to as the 'ESC Manual'), looking at the drivers and benefits that a guidance manual in this area may provide. (authors)

  4. Pollutant Assessments Group procedures manual: Volume 2, Technical procedures

    Energy Technology Data Exchange (ETDEWEB)

    1992-03-01

    This is volume 2 of the manuals that describes the technical procedures currently in use by the Pollution Assessments Group. This manual incorporates new developments in hazardous waste assessment technology and administrative policy. Descriptions of the equipment, procedures and operations of such things as radiation detection, soil sampling, radionuclide monitoring, and equipment decontamination are included in this manual. (MB)

  5. Occupational safety manual. 4. rev. ed.

    International Nuclear Information System (INIS)

    Skiba, R.

    1979-01-01

    With the rapid technical development and increasing life expectancy, protection against occupational hazards has become a central economic and social problem. Consultants and others responsible for occupational safety should not only be interested in their work but also have sufficient basic knowledge. The complex subject is presented in a simplified and easily comprehensible manner, with particular regard to the practical side. The book is at the same time a textbook for basic training and a reference manual for the working practice. (orig./HP) [de

  6. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

    On June 27, 1989, the Secretary of Energy announced a 10-Point Initiative to strengthen environment, safety, and health (ES ampersand H) programs, and waste management activities at DOE production, research, and testing facilities. One of the points involved conducting dent Tiger Team Assessments of DOE operating facilities. The Office of Special independent Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This manual documents the processes to be used to perform the ES ampersand H Progress Assessments. It was developed based upon the lessons learned from Tiger Team Assessments, the two pilot Progress Assessments, and Progress Assessments that have been completed. The manual will be updated periodically to reflect lessons learned or changes in policy

  7. Unique safety manual for experimental personnel

    International Nuclear Information System (INIS)

    Busick, D.D.; Warren, G.J.

    1979-01-01

    Within a few months of the discovery of x-rays the first radiation injuries were reported (ta71). During the past thirty years both the number and complexity of x-ray analytical units have increased markedly. The world-wide number of incidents leading to severe injury has also increased. For analytical x-ray machines the need for engineered and administrative safeguards has long been recognized. At Stanford Synchrotron Radiation Laboratory (SSRL) the personnel protection system has been carefully designed to maximize safety and minimize experimental interference. However, all possible experimental configurations cannot be anticipated and some interference is to be expected. There are means by which safeguards can be substituted as long as these substitutions do not degrade the existing degree of safety. any substitutions must be evaluated by the Radiation Safety Committee, the SSRL staff and Operational Health Physics. Some studies have indicated that between fifty and ninety percent of serious radiation accidents are directly related to human errors, i.e., ignoring administrative proccedures, by-passing engineered safeguards or by inadequate training. Lindell has estimated the annual probability of serious injury to be about 1:100 per macchine. No matter what the real probability of serious injury is the personnel protection system should reduce this risk to a value that approaches zero. It is hoped that this manual will bring into sharper focus some of the more serious results of unnecessary risk taking. We also hope that it will convey the very real necessity for safeguards which may at times appear to be arbitrary and unnecessary impediments to experimental purposes

  8. RADTRAN 4: Volume 4, Programmer's manual

    International Nuclear Information System (INIS)

    Kanipe, F.L.; Neuhauser, K.S.

    1992-07-01

    The RADTRAN 4 computer code is designed to analyze radiological consequences and accident risks of transporting radioactive material. This manual provides information useful for interpreting, troubleshooting, or debugging components of the code during development or revision of the program

  9. Selection methodology for LWR safety R and D programs and proposals. Volume III. User's manual for the multi-attribute utility package (MAUP)

    International Nuclear Information System (INIS)

    Hale, M.; Turnage, J.J.; Husseiny, A.A.; Ritzman, R.L.

    1981-02-01

    The computer program which was developed to apply the multi-attribute utility (MAU) methodology to the selection of LWR safety R and D programs and proposals is described. An overview of the MAU method is presented, followed by a description of the steps incorporated in developing individual modules for use in the multi-attribute utility package (MAUP). Each module is described complete with usage information and an example of computer output

  10. MARS code manual volume II: input requirements

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This input manual provides a complete list of input required to run MARS. The manual is divided largely into two parts, namely, the one-dimensional part and the multi-dimensional part. The inputs for auxiliary parts such as minor edit requests and graph formatting inputs are shared by the two parts and as such mixed input is possible. The overall structure of the input is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  11. GO methodology. Volume 1. Overview manual

    International Nuclear Information System (INIS)

    1983-06-01

    The GO methodology is a success-oriented probabilistic system performance analysis technique. The methodology can be used to quantify system reliability and availability, identify and rank critical components and the contributors to system failure, construct event trees, and perform statistical uncertainty analysis. Additional capabilities of the method currently under development will enhance its use in evaluating the effects of external events and common cause failures on system performance. This Overview Manual provides a description of the GO Methodology, how it can be used, and benefits of using it in the analysis of complex systems

  12. Principles and Practices of Occupational Safety and Health: Administrator's Manual.

    Science.gov (United States)

    Occupational Safety and Health Administration, Washington, DC.

    The manual guides an instructor in conducting a training course for first-line supervisors to familiarize them with six aspects relating to the Occupational Safety and Health Act of 1970: (1) requirements of the Act, (2) compliance with its standards, (3) identification of health and safety hazards, (4) correction of adverse conditions, (5) record…

  13. Federal Radiological Monitoring and Assessment Center Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

    FRMAC Health and Safety Working Group

    2012-03-20

    This manual is a tool to provide information to all responders and emergency planners and is suggested as a starting point for all organizations that provide personnel/assets for radiological emergency response. It defines the safety requirements for the protection of all emergency responders. The intent is to comply with appropriate regulations or provide an equal level of protection when the situation makes it necessary to deviate. In the event a situation arises which is not addressed in the manual, an appropriate management-level expert will define alternate requirements based on the specifics of the emergency situation. This manual is not intended to pertain to the general public.

  14. Manual on maintenance of systems and components important to safety

    International Nuclear Information System (INIS)

    1986-01-01

    The Manual should serve as guidance at the plant management level for the maintenance of systems and components important to safety. It includes a detailed description of management systems, administrative controls and procedures. The Annexes contain examples of documents and practices adopted by Operating Organizations of some Member States. It is not the intention of this Manual to address the technical problem of how to maintain a particular component but rather to cover the programmatic aspects of maintenance. It also contains some aspects of surveillance and verification activities. The Manual makes only general statements about radiation protection provisions in connection with maintenance; detailed guidance can be found in other IAEA documents

  15. Health and Safety Procedures Manual for hazardous waste sites

    Energy Technology Data Exchange (ETDEWEB)

    Thate, J.E.

    1992-09-01

    The Oak Ridge National Laboratory Chemical Assessments Team (ORNL/CAT) has developed this Health and Safety Procedures Manual for the guidance, instruction, and protection of ORNL/CAT personnel expected to be involved in hazardous waste site assessments and remedial actions. This manual addresses general and site-specific concerns for protecting personnel, the general public, and the environment from any possible hazardous exposures. The components of this manual include: medical surveillance, guidance for determination and monitoring of hazards, personnel and training requirements, protective clothing and equipment requirements, procedures for controlling work functions, procedures for handling emergency response situations, decontamination procedures for personnel and equipment, associated legal requirements, and safe drilling practices.

  16. Health and Safety Audit Design Manual

    Energy Technology Data Exchange (ETDEWEB)

    Ternes, Mark P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Langley, Brandon R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Accawi, Gina K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Malhotra, Mini [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-10-01

    The Health and Safety Audit is an electronic audit tool developed by the Oak Ridge National Laboratory to assist in the identification and selection of health and safety measures when a home is being weatherized (i.e., receiving home energy upgrades), especially as part of the US Department of Energy (DOE) Weatherization Assistance Program, or during home energy-efficiency retrofit or remodeling jobs. The audit is specifically applicable to existing single-family homes (including mobile homes), and is generally applicable to individual dwelling units in low-rise multifamily buildings. The health and safety issues covered in the audit are grouped in nine categories: mold and moisture, lead, radon, asbestos, formaldehyde and volatile organic compounds (VOCs), combustion, pest infestation, safety, and ventilation. Development of the audit was supported by the US Department of Housing and Urban Development Office of Healthy Homes and Lead Hazard Control and the DOE Weatherization Assistance Program.

  17. Load research manual. Volume 1. Load research procedures

    Energy Technology Data Exchange (ETDEWEB)

    Brandenburg, L.; Clarkson, G.; Grund, Jr., C.; Leo, J.; Asbury, J.; Brandon-Brown, F.; Derderian, H.; Mueller, R.; Swaroop, R.

    1980-11-01

    This three-volume manual presents technical guidelines for electric utility load research. Special attention is given to issues raised by the load data reporting requirements of the Public Utility Regulatory Policies Act of 1978 and to problems faced by smaller utilities that are initiating load research programs. In Volumes 1 and 2, procedures are suggested for determining data requirements for load research, establishing the size and customer composition of a load survey sample, selecting and using equipment to record customer electricity usage, processing data tapes from the recording equipment, and analyzing the data. Statistical techniques used in customer sampling are discussed in detail. The costs of load research also are estimated, and ongoing load research programs at three utilities are described. The manual includes guides to load research literature and glossaries of load research and statistical terms.

  18. Nuclear criticality safety. Chapter 0530 of AEC manual

    International Nuclear Information System (INIS)

    2006-01-01

    The programme objectives of this chapter of the U.S. Atomic Energy Commission manual on nuclear criticality safety are to protect the health and safety of the public and of the government and contractor personnel working in plants that handle fissionable material and to protect public and private property from the consequences of a criticality accident occurring in AEC-owned plants and other AEC-contracted activities involving fissionable materials

  19. Tritium radioluminescent devices, Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information.

  20. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information

  1. INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

    International Nuclear Information System (INIS)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User's Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

  2. INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny; Edlund, O; Hermann, J; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User`s Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

  3. Helicopter Operations and Personnel Safety (Helirescue Manual). Fourth Edition.

    Science.gov (United States)

    Dalle-Molle, John

    The illustrated manual includes information on various aspects of helicopter rescue missions, including mission management roles for key personnel, safety rules around helicopters, requests for helicopter support, sample military air support forms, selection of landing zones, helicopter evacuations, rescuer delivery, passenger unloading, crash…

  4. ACARP Project C10059. ACARP manual of modern coal testing methods. Volume 1: The manual

    Energy Technology Data Exchange (ETDEWEB)

    Sakurovs, R.; Creelman, R.; Pohl, J.; Juniper, L. [CSIRO Energy Technology, Sydney, NSW (Australia)

    2002-07-01

    The Manual summarises the purpose, applicability, and limitations of a range of standard and modern coal testing methods that have potential to assist the coal company technologist to better evaluate coal performance. The first volume sets out the Modern Coal Testing Methods in summarised form that can be used as a quick guide to practitioners to assist in selecting the best technique to solve their problems.

  5. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  6. Synfuel program analysis. Volume 2: VENVAL users manual

    Science.gov (United States)

    Muddiman, J. B.; Whelan, J. W.

    1980-07-01

    This volume is intended for program analysts and is a users manual for the VENVAL model. It contains specific explanations as to input data requirements and programming procedures for the use of this model. VENVAL is a generalized computer program to aid in evaluation of prospective private sector production ventures. The program can project interrelated values of installed capacity, production, sales revenue, operating costs, depreciation, investment, dent, earnings, taxes, return on investment, depletion, and cash flow measures. It can also compute related public sector and other external costs and revenues if unit costs are furnished.

  7. TV Trouble-Shooting Manual. Volumes 7-8. Part 3: Synchronisation and Deflection Circuits. Student and Instructor's Manuals.

    Science.gov (United States)

    Mukai, Masaaki; Kobayashi, Ryozo

    These volumes are, respectively, the self-instructional student manual and the teacher manual that cover the third set of training topics in this course for television repair technicians. Both contain identical information on synchronization and deflection circuits, including sections on the principle of synchronized deflection, synchronization…

  8. A feasibility study for Arizona's roadway safety management process using the Highway Safety Manual and SafetyAnalyst : final report.

    Science.gov (United States)

    2016-07-01

    To enable implementation of the American Association of State Highway Transportation (AASHTO) Highway Safety Manual using : SaftetyAnalyst (an AASHTOWare software product), the Arizona Department of Transportation (ADOT) studied the data assessment :...

  9. Quality assurance program manual for nuclear power plants. Volume I. Policies

    International Nuclear Information System (INIS)

    1976-01-01

    The Consumers Power Company Quality Assurance Program Manual for Nuclear Power Plants consists of policies and procedures which comply with current NRC regulatory requirements and industry codes and standards in effect during the design, procurement, construction, testing, operation, refueling, maintenance, repair and modification activities associated with nuclear power plants. Specific NRC and industry documents that contain the requirements, including the issue dates in effect, are identified in each nuclear power plant's Safety Analysis Report. The requirements established by these documents form the basis for the Consumer Power Quality Assurance Program, which is implemented to control those structures, systems, components and operational safety actions listed in each nuclear power plant's Quality List (Q-List). As additional and revised requirements are issued by the NRC and professional organizations involved in nuclear activities, they will be reviewed for their impact on this manual, and changes will be made where considered necessary. CP Co 1--Consumers Power Company QA Program Topical Report is Volume I of this manual and contains Quality Assurance Program Policies applicable during all phases of nuclear power plant design, construction and operation

  10. Manual on oil-gas industry waste utilization radioecological safety

    International Nuclear Information System (INIS)

    Kudryashev, V.A.; Lukashenko, S.N.; Tuleushev, A.Zh.; Marabaev, Zh.N.; Pasysaev, V.A.; Kayukov, P.G.; Kozhakhmetov, N.B.; Shevtsov, S.P.

    2003-01-01

    The development of a new document - 'Manual on radio-ecologically safe utilization of waste from oil-and-gas production' is carried out. This document regulates the whole cycle of environment protection measures at waste utilization for the named industry in Kazakhstan and is aimed on lowering the radiation risks and assurance of radioecological safety both at present and for the future. The document presents a set regulations necessary for radioactive wastes handling in the oil-gas industry. The normative document was agreed in both the Ministry of Health of the Republic of Kazakhstan (RK) and Ministry of Environment Protection of RK

  11. Safety handling manual for high dose rate remote afterloading system

    International Nuclear Information System (INIS)

    1999-01-01

    This manual is mainly for safety handling of 192 Ir-RALS (remote afterloading system) of high dose rate and followings were presented: Procedure and document format for the RALS therapy and for handling of its radiation source with the purpose of prevention of human errors and unexpected accidents, Procedure for preventing errors occurring in the treatment schedule and operation, and Procedure and format necessary for newly introducing the system into a facility. Consistency was intended in the description with the quality assurance guideline for therapy with small sealed radiation sources made by JASTRO (Japan Society for Therapeutic Radiology and Oncology). Use of the old type 60 Co-RALS was pointed out to be a serious problem remained and its safety handling procedure was also presented. (K.H.)

  12. Fire and blast safety manual for fuel element manufacture

    International Nuclear Information System (INIS)

    Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

    1988-01-01

    The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO 2 , MOX and HTR fuel elements. (orig./DG) [de

  13. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

  14. The Radiological Safety Analysis Computer Program (RSAC-5) user's manual

    International Nuclear Information System (INIS)

    Wenzel, D.R.

    1994-02-01

    The Radiological Safety Analysis Computer Program (RSAC-5) calculates the consequences of the release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory from either reactor operating history or nuclear criticalities. RSAC-5 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated through the inhalation, immersion, ground surface, and ingestion pathways. RSAC+, a menu-driven companion program to RSAC-5, assists users in creating and running RSAC-5 input files. This user's manual contains the mathematical models and operating instructions for RSAC-5 and RSAC+. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-5 and RSAC+. These programs are designed for users who are familiar with radiological dose assessment methods

  15. Principles and Practices of Occupational Safety and Health: Student Manual: Booklet One.

    Science.gov (United States)

    Occupational Safety and Health Administration, Washington, DC.

    The manual is the first of six student manuals for use in a course on occupational health and safety for supervisory personnel. The manual contains lessons 1-3 of the 15 consecutively-numbered lessons, each of which contains study questions (and answers) interwoven with the text and review questions at the end of each section. Lesson 1 (three…

  16. Principles and Practices of Occupational Safety and Health: Student Manual: Booklet Five.

    Science.gov (United States)

    Occupational Safety and Health Administration, Washington, DC.

    The manual is the fifth of six student manuals for use in a course on occupational health and safety for supervisory personnel. The manual contains lessons 14 and 15 of the 15 consecutively-numbered lessons, each of which contains study questions (and answers) interwoven with the text and review questions at the end of each section. Lesson 14…

  17. Small Business Management. Instructor's Manual. Volume I. Third Edition.

    Science.gov (United States)

    Jeanneau, Joseph A.; And Others

    The instructor's manual is one of four prepared as a guide in conducting a small Business Management course for American Indians to prepare them for jobs as owners/managers of their own businesses and for management positions with business owned by bonds, cooperatives, and others. The manual contains lesson plans, suggested methodologies, and…

  18. Safety manual for civil engineering and building works of nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    This manual lays down the various important considerations that go into safe design of civil structures for nuclear power plants. This manual identifies the design approach, quality assurance requirements and acceptance criteria that need to be observed to assure safety. Considerations on civil design having bearing on safety, during decommissioning are also indicated. (original). 37 refs., tabs

  19. ACARP Project C10059. ACARP manual of modern coal testing methods. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Sakurovs, R.; Creelman, R.; Pohl, J.; Juniper, L. [CSIRO Energy Technology, Sydney, NSW (Australia)

    2002-07-01

    The Manual summarises the purpose, applicability, and limitations of a range of standard and modern coal testing methods that have potential to assist the coal company technologist to better evaluate coal performance. It is presented in two volumes. This second volume provides more detailed information regarding the methods discussed in Volume 1.

  20. RADTRAN 4: Volume 4, Programmer`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Kanipe, F L [GRAM, Inc., Albuquerque, NM (United States); Neuhauser, K S [Sandia National Labs., Albuquerque, NM (United States)

    1992-07-01

    The RADTRAN 4 computer code is designed to analyze radiological consequences and accident risks of transporting radioactive material. This manual provides information useful for interpreting, troubleshooting, or debugging components of the code during development or revision of the program.

  1. Stepping motor control processor reference manual. Volume I

    International Nuclear Information System (INIS)

    Holloway, F.W.; VanArsdall, P.J.; Suski, G.J.; Gant, R.G.; Rash, M.

    1980-01-01

    This manual is intended to serve several purposes. The first goal is to describe the capabilities and operation of the SMC processor package from an operator or user point of view. Secondly, the manual will describe in some detail the basic hardware elements and how they can be used effectively to implement a step motor control system. Practical information on the use, installation and checkout of the hardware set is presented in the following sections along with programming suggestions. Available related system software is described in this manual for reference and as an aid in understanding the system architecture. Section two presents an overview and operations manual of the SMC processor describing its composition and functional capabilities. Section three contains hardware descriptions in some detail for the LLL-designed hardware used in the SMC processor. Basic theory of operation and important features are explained

  2. Federal Radiological Monitoring and Assessment Center Monitoring Manual Volume 2, Radiation Monitoring and Sampling

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Aerial Measurement Systems

    2012-07-31

    The FRMAC Monitoring and Sampling Manual, Volume 2 provides standard operating procedures (SOPs) for field radiation monitoring and sample collection activities that are performed by the Monitoring group during a FRMAC response to a radiological emergency.

  3. Manual of functions, assignments, and responsibilities for nuclear safety: Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-10-15

    The FAR Manual is a convenient easy-to-use collection of the functions, assignments, and responsibilities (FARs) of DOE nuclear safety personnel. Current DOE directives, including Orders, Secretary of Energy Notices, and other assorted policy memoranda, are the source of this information and form the basis of the FAR Manual. Today, the majority of FARs for DOE personnel are contained in DOE`s nuclear safety Orders. As these Orders are converted to rules in the Code of Federal Regulations, the FAR Manual will become the sole source for information relating to the functions, assignments, responsibilities of DOE nuclear safety personnel. The FAR Manual identifies DOE directives that relate to nuclear safety and the specific DOE personnel who are responsible for implementing them. The manual includes only FARs that have been extracted from active directives that have been approved in accordance with the procedures contained in DOE Order 1321.1B.

  4. Laboratory Safety Manual for Alabama Schools. Bulletin 1975. No. 20.

    Science.gov (United States)

    Alabama State Dept. of Education, Montgomery.

    This document presents the Alabama State Department of Education guidelines for science laboratory safety, equipment, storage, chemical safety, rocket safety, electrical safety, safety with radioisotopes, and safety with biologicals. Also included is a brief bibliography, a teacher's checklist, a listing of laser facts and regulations, and a…

  5. MARS CODE MANUAL VOLUME V: Models and Correlations

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Bae, Sung Won; Lee, Seung Wook; Yoon, Churl; Hwang, Moon Kyu; Kim, Kyung Doo; Jeong, Jae Jun

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  6. Impacts of Outer Continental Shelf (OCS) development on recreation and tourism. Volume 5. Program logic manual

    Energy Technology Data Exchange (ETDEWEB)

    1987-04-01

    The final report for the project is presented in five volumes. This volume is the Programmer's Manual. It covers: a system overview, attractiveness component of gravity model, trip-distribution component of gravity model, economic-effects model, and the consumer-surplus model. The project sought to determine the impact of Outer Continental Shelf development on recreation and tourism.

  7. Highway Safety Manual applied in Missouri - freeway/software : research summary.

    Science.gov (United States)

    2016-03-01

    AASHTOs Highway Safety Manual (HSM) : includes models for freeway segments, speedchange : lanes (transitional area between mainline : and ramps), ramps, and interchange terminals. : These predictive models for freeway : interchanges need to be cal...

  8. Technical specifications manual for the MARK-1 pulsed ionizing radiation detection system. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, R.S.; Harker, Y.D.; Jones, J.L.; Hoggan, J.M.

    1993-03-01

    The MARK-1 detection system was developed by the Idaho National Engineering Laboratory for the US Department of Energy Office of Arms Control and Nonproliferation. The completely portable system was designed for the detection and analysis of intense photon emissions from pulsed ionizing radiation sources. This manual presents the technical design specifications for the MARK-1 detection system and was written primarily to assist the support or service technician in the service, calibration, and repair of the system. The manual presents the general detection system theory, the MARK-1 component design specifications, the acquisition and control software, the data processing sequence, and the system calibration procedure. A second manual entitled: Volume 2: Operations Manual for the MARK-1 Pulsed Ionizing Radiation Detection System (USDOE Report WINCO-1108, September 1992) provides a general operational description of the MARK-1 detection system. The Operations Manual was written primarily to assist the field operator in system operations and analysis of the data.

  9. Fixed site neutralization model programmer's manual. Volume II

    International Nuclear Information System (INIS)

    Engi, D.; Chapman, L.D.; Judnick, W.; Blum, R.; Broegler, L.; Lenz, J.; Weinthraub, A.; Ballard, D.

    1979-12-01

    This report relates to protection of nuclear materials at nuclear facilities. This volume presents the source listings for the Fixed Site Neutralization Model and its supporting modules, the Plex Preprocessor and the Data Preprocessor

  10. Manual on quality assurance for computer software related to the safety of nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    The objective of the Manual is to provide guidance in the assurance of quality of specification, design, maintenance and use of computer software related to items and activities important to safety (hereinafter referred to as safety related) in nuclear power plants. This guidance is consistent with, and supplements, the requirements and recommendations of Quality Assurance for Safety in Nuclear Power Plants: A Code of Practice, 50-C-QA, and related Safety Guides on quality assurance for nuclear power plants. Annex A identifies the IAEA documents referenced in the Manual. The Manual is intended to be of use to all those who, in any way, are involved with software for safety related applications for nuclear power plants, including auditors who may be called upon to audit management systems and product software. Figs

  11. Stimulation model for lenticular sands: Volume 2, Users manual

    Energy Technology Data Exchange (ETDEWEB)

    Rybicki, E.F.; Luiskutty, C.T.; Sutrick, J.S.; Palmer, I.D.; Shah, G.H.; Tomutsa, L.

    1987-07-01

    This User's Manual contains information for four fracture/proppant models. TUPROP1 contains a Geertsma and de Klerk type fracture model. The section of the program utilizing the proppant fracture geometry data from the pseudo three-dimensional highly elongated fracture model is called TUPROPC. The analogous proppant section of the program that was modified to accept fracture shape data from SA3DFRAC is called TUPROPS. TUPROPS also includes fracture closure. Finally there is the penny fracture and its proppant model, PENNPROP. In the first three chapters, the proppant sections are based on the same theory for determining the proppant distribution but have modifications to support variable height fractures and modifications to accept fracture geometry from three different fracture models. Thus, information about each proppant model in the User's Manual builds on information supplied in the previous chapter. The exception to the development of combined treatment models is the penny fracture and its proppant model. In this case, a completely new proppant model was developed. A description of how to use the combined treatment model for the penny fracture is contained in Chapter 4. 2 refs.

  12. Considerations to improve the safety of cervical spine manual therapy.

    Science.gov (United States)

    Hutting, Nathan; Kerry, Roger; Coppieters, Michel W; Scholten-Peeters, Gwendolijne G M

    2018-02-01

    Manipulation and mobilisation of the cervical spine are well established interventions in the management of patients with headache and/or neck pain. However, their benefits are accompanied by potential, yet rare risks in terms of serious adverse events, including neurovascular insult to the brain. A recent international framework for risk assessment and management offers directions in the mitigation of this risk by facilitating sound clinical reasoning. The aim of this article is to critically reflect on and summarize the current knowledge about cervical spine manual therapy and to provide guidance for clinical reasoning for cervical spine manual therapy. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Space Propulsion Hazards Analysis Manual (SPHAM). Volume 2. Appendices

    Science.gov (United States)

    1988-10-01

    Vanderwall, E.M. and Schaplowsky, R.F., USAF PROPELLANT HA& nBOOKS , Volume III, Part B, "Nitrogen Trifluoride, Bibliography," p. 4-508, Aerojeft Liquid...not economically desirable to maintain two different avionic configurations in the space-based program. Guidance and navigation information is

  14. Light Water Reactor Generic Safety Issues Database (LWRGSIDB). User's manual

    International Nuclear Information System (INIS)

    1999-01-01

    The IAEA Conference on 'The Safety of Nuclear Power: Strategy for the Future' in 1991 was a milestone in nuclear safety. The objective of this conference was to review nuclear power safety issues for which achieving international consensus would be desirable, to address concerns on nuclear safety and to formulate recommendations for future actions by national and international authorities to advance nuclear safety to the highest level. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. Some of the publications related to these two items that have been issued subsequent to this conference are: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-8 (1995), the IAEA Safety Guide 50-SG-O12, Periodic Safety Review of Operational Nuclear Power Plants (1994) and IAEA Safety Reports Series No. 12, Evaluation of the Safety of Operating Nuclear Power Plants Built to Earlier Standards: A Common Basis for Judgement (1998). Some of the findings of the 1991 conference have not yet been fully addressed. An IAEA Symposium on Reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review those operating nuclear power plants which do not reach the high safety levels of the vast majority of plants and to undertake improvements with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear Safety and its implementation. In order to perform safety reviews and to reassess the safety of operating nuclear power plants in a uniform manner, it is imperative to have an internationally accepted reference. Existing guidance needs to be complemented by a list of safety issues which have been encountered and

  15. ABAQUS-EPGEN: a general-purpose finite element code. Volume 3. Example problems manual

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.

    1983-03-01

    This volume is the Example and Verification Problems Manual for ABAQUS/EPGEN. Companion volumes are the User's, Theory and Systems Manuals. This volume contains two major parts. The bulk of the manual (Sections 1-8) contains worked examples that are discussed in detail, while Appendix A documents a large set of basic verification cases that provide the fundamental check of the elements in the code. The examples in Sections 1-8 illustrate and verify significant aspects of the program's capability. Most of these problems provide verification, but they have also been chosen to allow discussion of modeling and analysis techniques. Appendix A contains basic verification cases. Each of these cases verifies one element in the program's library. The verification consists of applying all possible load or flux types (including thermal loading of stress elements), and all possible foundation or film/radiation conditions, and checking the resulting force and stress solutions or flux and temperature results. This manual provides program verification. All of the problems described in the manual are run and the results checked, for each release of the program, and these verification results are made available

  16. 76 FR 60511 - Amendment of Marine Safety Manual, Volume III

    Science.gov (United States)

    2011-09-29

    ...'' in the ``Keyword'' box. Click ``Search,'' and then click on the balloon shape in the ``Actions... Comments'' box, which will then become highlighted in blue. In the ``Keyword'' box, insert ``USCG- 2011...

  17. Field Operations and Enforcement Manual for Air Pollution Control. Volume III: Inspection Procedures for Specific Industries.

    Science.gov (United States)

    Weisburd, Melvin I.

    The Field Operations and Enforcement Manual for Air Pollution Control, Volume III, explains in detail the following: inspection procedures for specific sources, kraft pulp mills, animal rendering, steel mill furnaces, coking operations, petroleum refineries, chemical plants, non-ferrous smelting and refining, foundries, cement plants, aluminum…

  18. The Gift Code User Manual. Volume I. Introduction and Input Requirements

    Science.gov (United States)

    1975-07-01

    REPORT & PERIOD COVERED ‘TII~ GIFT CODE USER MANUAL; VOLUME 1. INTRODUCTION AND INPUT REQUIREMENTS FINAL 6. PERFORMING ORG. REPORT NUMBER ?. AuTHOR(#) 8...reverua side if neceaeary and identify by block number] (k St) The GIFT code is a FORTRANcomputerprogram. The basic input to the GIFT ode is data called

  19. Different approaches to synovial membrane volume determination by magnetic resonance imaging: manual versus automated segmentation

    DEFF Research Database (Denmark)

    Østergaard, Mikkel

    1997-01-01

    Automated fast (5-20 min) synovial membrane volume determination by MRI, based on pre-set post-gadolinium-DTPA enhancement thresholds, was evaluated as a substitute for a time-consuming (45-120 min), previously validated, manual segmentation method. Twenty-nine knees [rheumatoid arthritis (RA) 13...

  20. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1985-07-01

    The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

  1. MELCOR computer code manuals: Primer and user's guides, Version 1.8.3 September 1994. Volume 1

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users' Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  2. Safety Specialist Manpower, Manpower Resources. Volumes II and III.

    Science.gov (United States)

    Booz Allen and Hamilton, Inc., Washington, DC.

    These second and third volumes of a four-volume study of manpower in state highway safety programs over the next decade estimate manpower resources by state and in national aggregate and describe present and planned training programs for safety specialists. For each educational level, both total manpower and manpower actually available for…

  3. Fixed site neutralization model programmer's manual. Volume 1

    International Nuclear Information System (INIS)

    Engi, D.; Chapman, L.; Judnick, W.; Blum, R.; Broegler, L.; Lenz, J.; Weinthraub, A.; Ballard, D.

    1979-12-01

    The Fixed Site Neutralization Model (FSNM) is a stochastic, time-stepped simulation of an engagement process whereby an adversary force attempts to steal or sabotage sensitive (e.g., nuclear) materials being guarded by a security force on a fixed site and a response force that is off-site. FSNM will assist regulatory bodies of the US Government in evaluating fixed site physical protection systems at various installations in a variety of scenarios. Most data input to the Fixed Site Neutralization Model are in binary form. The user's data are transformed into binary form by two supporting modules, the Data Preprocessor (DPP) and the Plex Preprocessor (PPP). Both preprocessors and the FSNM itself are written in FORTRAN. This volume of the program maintenance program contains: introduction, logic descriptions, PLEX data structure, PLEX records, dictionaries, and error conditions and comments

  4. Development of a health and safety manual for emergency response operations

    International Nuclear Information System (INIS)

    Riland, C.A.; Junio, S.S.

    2000-01-01

    The Federal Radiological Monitoring and Assessment Center (FRMAC) Health and Safety Manual, which has been under development by a multi-agency group, is nearing completion and publication. The manual applies to offsite monitoring during a radiological accident or incident. Though written for multi-agency offsite monitoring activities (FRMAC), the manual is generic in nature and should be readily adaptable for other emergency response operations. Health and safety issues for emergency response situations often differ from those of normal operations. Examples of these differences and methodologies to address these issues are discussed. Challenges in manual development, including lack of regulatory and guidance documentation, are also discussed. One overriding principle in the Health and Safety Manual development is the overall reduction of risk, not just dose. The manual is broken into several chapters, which include Overview of Responsibities, Health Physics, Industrial Hygiene and Safey, Medical, and Environmental Compliance and Records. Included is a series of appendices, which presents additional information on forms and plans for default scenarios

  5. GASFLOW: A Computational Fluid Dynamics Code for Gases, Aerosols, and Combustion, Volume 2: User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, B. D.; Mueller, C.; Necker, G. A.; Travis, J. R.; Spore, J. W.; Lam, K. L.; Royl, P.; Wilson, T. L.

    1998-10-01

    Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume III

  6. Load research manual. Volume 2. Fundamentals of implementing load research procedures

    Energy Technology Data Exchange (ETDEWEB)

    Brandenburg, L.; Clarkson, G.; Grund, Jr., C.; Leo, J.; Asbury, J.; Brandon-Brown, F.; Derderian, H.; Mueller, R.; Swaroop, R.

    1980-11-01

    This three-volume manual presents technical guidelines for electric utility load research. Special attention is given to issues raised by the load data reporting requirements of the Public Utility Regulatory Policies Act of 1978 and to problems faced by smaller utilities that are initiating load research programs. In Volumes 1 and 2, procedures are suggested for determining data requirements for load research, establishing the size and customer composition of a load survey sample, selecting and using equipment to record customer electricity usage, processing data tapes from the recording equipment, and analyzing the data. Statistical techniques used in customer sampling are discussed in detail. The costs of load research also are estimated, and ongoing load research programs at three utilities are described. The manual includes guides to load research literature and glossaries of load research and statistical terms.

  7. Valid and efficient manual estimates of intracranial volume from magnetic resonance images

    International Nuclear Information System (INIS)

    Klasson, Niklas; Olsson, Erik; Rudemo, Mats; Eckerström, Carl; Malmgren, Helge; Wallin, Anders

    2015-01-01

    Manual segmentations of the whole intracranial vault in high-resolution magnetic resonance images are often regarded as very time-consuming. Therefore it is common to only segment a few linearly spaced intracranial areas to estimate the whole volume. The purpose of the present study was to evaluate how the validity of intracranial volume estimates is affected by the chosen interpolation method, orientation of the intracranial areas and the linear spacing between them. Intracranial volumes were manually segmented on 62 participants from the Gothenburg MCI study using 1.5 T, T 1 -weighted magnetic resonance images. Estimates of the intracranial volumes were then derived using subsamples of linearly spaced coronal, sagittal or transversal intracranial areas from the same volumes. The subsamples of intracranial areas were interpolated into volume estimates by three different interpolation methods. The linear spacing between the intracranial areas ranged from 2 to 50 mm and the validity of the estimates was determined by comparison with the entire intracranial volumes. A progressive decrease in intra-class correlation and an increase in percentage error could be seen with increased linear spacing between intracranial areas. With small linear spacing (≤15 mm), orientation of the intracranial areas and interpolation method had negligible effects on the validity. With larger linear spacing, the best validity was achieved using cubic spline interpolation with either coronal or sagittal intracranial areas. Even at a linear spacing of 50 mm, cubic spline interpolation on either coronal or sagittal intracranial areas had a mean absolute agreement intra-class correlation with the entire intracranial volumes above 0.97. Cubic spline interpolation in combination with linearly spaced sagittal or coronal intracranial areas overall resulted in the most valid and robust estimates of intracranial volume. Using this method, valid ICV estimates could be obtained in less than five

  8. LLCEDATA and LLCECALC for Windows version 1.0, Volume 2: Technical manual

    Energy Technology Data Exchange (ETDEWEB)

    McFadden, J.G.

    1998-09-04

    LLCEDATA and LLCECALC for Windows are user-friendly computer software programs that work together to determine the proper waste designation, handling, and disposition requirements for Long Length Contaminated Equipment (LLCE). LLCEDATA reads from a variety of databases to produce an equipment data file (EDF) that represents a snapshot of both the LLCE and the tank from which it originates. LLCECALC reads the EDF and the gamma assay (AV2) file that is produced by the Flexible Receiver Gamma Energy Analysis System. LLCECALC performs corrections to the AV2 file as it is being read and characterizes the LLCE Both programs produce a variety of reports, including a characterization report and a status report. The status report documents each action taken by the user, LLCEDA TA, and LLCECALC. Documentation for LLCEDA TA and LLCECALC for Windows is available in three volumes. Volume 1 is a user`s manual, which is intended as a quick reference for both LLCEDATA and LLCECALC. Volume 2 is a technical manual, which documents LLCEDATA and LLCECALCs methodology, structure, and format, provides sample calculations; discusses system limitations; and provides recommendations to the LLCE process. Volume 3 is a software verification and validation document LLCEDATA and LLCECALC for Windows are sponsored by the U.S.Department of Energy. They were developed and are maintained by Engineering, Waste Management Federal Services, Inc., Northwest Operations. Copies of the software and documentation are available by contacting Engineering, Waste Management Federal Services, Inc., Northwest Operations, Richland, Washington.

  9. LLCEDATA and LLCECALC for Windows version 1.0, Volume 1: User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    McFadden, J.G.

    1998-09-04

    LLCEDATA and LLCECALC for Windows are user-friendly computer software programs that work together to determine the proper waste designation, handling, and disposition requirements for Long Length Contaminated Equipment (LLCE). LLCEDATA reads from a variety of data bases to produce an equipment data file (EDF) that represents a snapshot of both the LLCE and the tank it originates from. LLCECALC reads the EDF and a gamma assay (AV2) file that is produced by the Flexible Receiver Gamma Energy Analysis System. LLCECALC performs corrections to the AV2 file as it is being read and characterizes the LLCE. Both programs produce a variety of reports, including a characterization report and a status report. The status report documents each action taken by the user, LLCEDATA, and LLCECALC. Documentation for LLCEDATA and LLCECALC for Windows is available in three volumes. Volume 1 is a user`s manual, which is intended as a quick reference for both LLCEDATA and LLCECALC. Volume 2 is a technical manual, and Volume 3 is a software verification and validation document.

  10. LLCEDATA and LLCECALC for Windows version 1.0, Volume 1: User's manual

    International Nuclear Information System (INIS)

    McFadden, J.G.

    1998-01-01

    LLCEDATA and LLCECALC for Windows are user-friendly computer software programs that work together to determine the proper waste designation, handling, and disposition requirements for Long Length Contaminated Equipment (LLCE). LLCEDATA reads from a variety of data bases to produce an equipment data file (EDF) that represents a snapshot of both the LLCE and the tank it originates from. LLCECALC reads the EDF and a gamma assay (AV2) file that is produced by the Flexible Receiver Gamma Energy Analysis System. LLCECALC performs corrections to the AV2 file as it is being read and characterizes the LLCE. Both programs produce a variety of reports, including a characterization report and a status report. The status report documents each action taken by the user, LLCEDATA, and LLCECALC. Documentation for LLCEDATA and LLCECALC for Windows is available in three volumes. Volume 1 is a user's manual, which is intended as a quick reference for both LLCEDATA and LLCECALC. Volume 2 is a technical manual, and Volume 3 is a software verification and validation document

  11. Manual on radiological safety in uranium and thorium mines and mills

    International Nuclear Information System (INIS)

    1976-01-01

    The manual describes the personnel radiation hazards in uranium and thorium mines and mills. Measures which should be taken in order to protect the workers are outlined. The problems of air born radioactivity, external radiation, surface contamination and radioactive waste are treated. Safety standards in relation to the above mentioned subjects are given. An outline is given for monitoring programme. Monitoring methods, control methods and means of medical control are given

  12. Vault Safety and Inventory System users manual, PRIME 2350. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Downey, N.J.

    1994-12-14

    This revision is issued to request review of the attached document: VSIS User Manual, PRIME 2350, which provides user information for the operation of the VSIS (Vault Safety and Inventory System). It describes operational aspects of Prime 2350 minicomputer and vault data acquisition equipment. It also describes the User`s Main Menu and menu functions, including REPORTS. Also, system procedures for the Prime 2350 minicomputer are covered.

  13. Vault Safety and Inventory System users manual, PRIME 2350. Revision 1

    International Nuclear Information System (INIS)

    Downey, N.J.

    1994-01-01

    This revision is issued to request review of the attached document: VSIS User Manual, PRIME 2350, which provides user information for the operation of the VSIS (Vault Safety and Inventory System). It describes operational aspects of Prime 2350 minicomputer and vault data acquisition equipment. It also describes the User's Main Menu and menu functions, including REPORTS. Also, system procedures for the Prime 2350 minicomputer are covered

  14. Abdominal fat volume estimation by stereology on CT: a comparison with manual planimetry

    Energy Technology Data Exchange (ETDEWEB)

    Manios, G.E.; Mazonakis, M.; Damilakis, J. [University of Crete, Department of Medical Physics, Faculty of Medicine, Heraklion, Crete (Greece); Voulgaris, C.; Karantanas, A. [University of Crete, Department of Radiology, Faculty of Medicine, Heraklion, Crete (Greece)

    2016-03-15

    To deploy and evaluate a stereological point-counting technique on abdominal CT for the estimation of visceral (VAF) and subcutaneous abdominal fat (SAF) volumes. Stereological volume estimations based on point counting and systematic sampling were performed on images from 14 consecutive patients who had undergone abdominal CT. For the optimization of the method, five sampling intensities in combination with 100 and 200 points were tested. The optimum stereological measurements were compared with VAF and SAF volumes derived by the standard technique of manual planimetry on the same scans. Optimization analysis showed that the selection of 200 points along with the sampling intensity 1/8 provided efficient volume estimations in less than 4 min for VAF and SAF together. The optimized stereology showed strong correlation with planimetry (VAF: r = 0.98; SAF: r = 0.98). No statistical differences were found between the two methods (VAF: P = 0.81; SAF: P = 0.83). The 95 % limits of agreement were also acceptable (VAF: -16.5 %, 16.1 %; SAF: -10.8 %, 10.7 %) and the repeatability of stereology was good (VAF: CV = 4.5 %, SAF: CV = 3.2 %). Stereology may be successfully applied to CT images for the efficient estimation of abdominal fat volume and may constitute a good alternative to the conventional planimetric technique. (orig.)

  15. Simulation model for wind energy storage systems. Volume II. Operation manual. [SIMWEST code

    Energy Technology Data Exchange (ETDEWEB)

    Warren, A.W.; Edsinger, R.W.; Burroughs, J.D.

    1977-08-01

    The effort developed a comprehensive computer program for the modeling of wind energy/storage systems utilizing any combination of five types of storage (pumped hydro, battery, thermal, flywheel and pneumatic). An acronym for the program is SIMWEST (Simulation Model for Wind Energy Storage). The level of detail of SIMWEST is consistent with a role of evaluating the economic feasibility as well as the general performance of wind energy systems. The software package consists of two basic programs and a library of system, environmental, and load components. Volume II, the SIMWEST operation manual, describes the usage of the SIMWEST program, the design of the library components, and a number of simple example simulations intended to familiarize the user with the program's operation. Volume II also contains a listing of each SIMWEST library subroutine.

  16. Quality assurance program manual for nuclear power plants. Volume I. Policies

    International Nuclear Information System (INIS)

    1975-01-01

    Policies and procedures are presented which comply with current NRC regulatory requirements and industry codes and standards in effect during the design, procurement, construction, testing, operation, refueling, maintenance, repair and modification activities associated with nuclear power plants. Specific NRC and industry documents that contain the requirements, including the issue dates in effect, are identified in each nuclear power plant's Safety Analysis Report. The requirements established by these documents form the basis for the Consumers Power Quality Assurance Program, which is implemented to control those structures, systems, components and operational safety actions listed in each nuclear power plant's Quality List (Q-List). As additional and revised requirements are issued by the NRC and professional organizations involved in nuclear activities, they will be reviewed for their impact on this manual, and changes will be made where considered necessary

  17. Abdominal fat volume estimation by stereology on CT: a comparison with manual planimetry.

    Science.gov (United States)

    Manios, G E; Mazonakis, M; Voulgaris, C; Karantanas, A; Damilakis, J

    2016-03-01

    To deploy and evaluate a stereological point-counting technique on abdominal CT for the estimation of visceral (VAF) and subcutaneous abdominal fat (SAF) volumes. Stereological volume estimations based on point counting and systematic sampling were performed on images from 14 consecutive patients who had undergone abdominal CT. For the optimization of the method, five sampling intensities in combination with 100 and 200 points were tested. The optimum stereological measurements were compared with VAF and SAF volumes derived by the standard technique of manual planimetry on the same scans. Optimization analysis showed that the selection of 200 points along with the sampling intensity 1/8 provided efficient volume estimations in less than 4 min for VAF and SAF together. The optimized stereology showed strong correlation with planimetry (VAF: r = 0.98; SAF: r = 0.98). No statistical differences were found between the two methods (VAF: P = 0.81; SAF: P = 0.83). The 95% limits of agreement were also acceptable (VAF: -16.5%, 16.1%; SAF: -10.8%, 10.7%) and the repeatability of stereology was good (VAF: CV = 4.5%, SAF: CV = 3.2%). Stereology may be successfully applied to CT images for the efficient estimation of abdominal fat volume and may constitute a good alternative to the conventional planimetric technique. Abdominal obesity is associated with increased risk of disease and mortality. Stereology may quantify visceral and subcutaneous abdominal fat accurately and consistently. The application of stereology to estimating abdominal volume fat reduces processing time. Stereology is an efficient alternative method for estimating abdominal fat volume.

  18. NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

    Science.gov (United States)

    Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

    2015-01-01

    This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.

  19. Thirteenth water reactor safety research information meeting: proceedings Volume 1

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1986-02-01

    This six-volume report contains 151 papers out of the 178 that were presented at the Thirteenth Water Reactor Safety Research Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 22-25, 1985. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-one different papers presented by researchers from Japan, Canada and eight European countries. The title of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume presents information on: risk analysis PRA application; severe accident sequence analysis; risk analysis/dependent failure analysis; and industry safety research

  20. Safety of magnetic fusion facilities: Volume 2, Guidance

    International Nuclear Information System (INIS)

    1995-01-01

    This document provides guidance for the implementation of the requirements identified in Vol. 1 of this Standard. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While Vol. 1 is generally applicable in that requirements there apply to a wide range of fusion facilities, this volume is concerned mainly with large facilities such as the International Thermonuclear Experimental Reactor (ITER). Using a risk-based prioritization, the concepts presented here may also be applied to other magnetic fusion facilities. This volume is oriented toward regulation in the Department of Energy (DOE) environment

  1. Manual on safety aspects of the design and equipment of hot laboratories. 1981 ed

    International Nuclear Information System (INIS)

    1981-01-01

    This manual covers the general principles of planning and design of areas inside laboratories according to the varying potential radiation and contamination hazards; enclosures for radioactive material containment; viewing and lighting systems and various types of manipulators; transfer and transport of radioactive materials within the laboratories; air cleaning and ventilation systems, with particular reference to IAEA Safety Series No.17; techniques for controlling air pollution from the operation of nuclear facilities; various radioactive waste disposal systems; criticality control; fire protection; personnel monitoring, including changing-room monitoring and protective clothing; standardization and automation; and administrative controls. Although alpha, beta, gamma technologies have developed separately, equipment used in radioactive work is common to many operations. There is a step change in technology between work with uranium and plutonium and between work with plutonium and other transuranics; with plutonium one enters the field of alpha, beta, gamma technology. This manual reports the basic requirements and gives reference to more sophisticated techniques available. It is not concerned with work on a commercial scale. Other publications of interest in this context, for instance IAEA Safety Series No.39, are referenced for more detailed information.

  2. PC based manual and safety logic card test setup for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Chandgadkar, G.M.; Kohli, A.K.; Agarwal, R.G.; Chandra, Rajesh

    1992-01-01

    Fuel handling controls for 235 MWe PHWR make use of Manual and Logic cards (MLCs) for providing safety interlocks. These cards consist of various type of logic blocks. By connecting these logic blocks all the safety interlocks required for fuel handling controls have been provided. Previously trouble shooting of these cards was done by means of logic probe. Since the method was manual, it was laborious and time consuming. PC based test setup has overcome this drawback and detects the fault at the component level within few seconds. It also gives printout of status of faulty MLC cards. Here motherboard has been designed having slots for insertion of MLC cards. The input/output connection of these cards are coming to two 50 pin FRC connectors. PC communicates through 144 line digital input/output card with MLC card under test. Software is user friendly and outputs suitable input patterns to the card under test and checks for output pattern. It compares this output pattern with compare pattern and detects the fault and displays the symptoms. This system is currently in use at test facility for fuelling machine for 235 MWe PHWR reactor at Refuelling Technology Division, Hall-7. This test setup has been proposed for use at NAPP and future reactors. (author). 4 figs., 1 annexure

  3. Relating safety, productivity and company type for motor-manual logging operations in the Italian Alps.

    Science.gov (United States)

    Montorselli, Niccolò Brachetti; Lombardini, Carolina; Magagnotti, Natascia; Marchi, Enrico; Neri, Francesco; Picchi, Gianni; Spinelli, Raffaele

    2010-11-01

    The study compared the performance of four different logging crews with respect to productivity, organization and safety. To this purpose, the authors developed a data collection method capable of providing a quantitative analysis of risk-taking behavior. Four crews were tested under the same working conditions, representative of close-to-nature alpine forestry. Motor-manual working methods were applied, since these methods are still prevalent in the specific study area, despite the growing popularity of mechanical processors. Crews from public companies showed a significantly lower frequency of risk-taking behavior. The best safety performance was offered by the only (public) crew that had been administered formal safety training. The study seems to deny the common prejudice that safety practice is inversely proportional to productivity. Instead, productivity is increased by introducing more efficient working methods and equipment. The quantitative analysis of risk-taking behavior developed in this study can be applied to a number of industrial fields besides forestry. Characterizing risk-taking behavior for a given case may eventually lead to the development of custom-made training programmes, which may address problem areas while avoiding that the message is weakened by the inclusion of redundant information. In the specific case of logging crews in the central Alps, the study suggests that current training courses may be weak on ergonomics, and advocates a staged training programme, focusing first on accident reduction and then expanding to the prevention of chronic illness. 2010 Elsevier Ltd. All rights reserved.

  4. Field Operations and Enforcement Manual for Air Pollution Control. Volume II: Control Technology and General Source Inspection.

    Science.gov (United States)

    Weisburd, Melvin I.

    The Field Operations and Enforcement Manual for Air Pollution Control, Volume II, explains in detail the following: technology of source control, modification of operations, particulate control equipment, sulfur dioxide removal systems for power plants, and control equipment for gases and vapors; inspection procedures for general sources, fuel…

  5. Automated and Semiautomated Segmentation of Rectal Tumor Volumes on Diffusion-Weighted MRI: Can It Replace Manual Volumetry?

    International Nuclear Information System (INIS)

    Heeswijk, Miriam M. van; Lambregts, Doenja M.J.; Griethuysen, Joost J.M. van; Oei, Stanley; Rao, Sheng-Xiang; Graaff, Carla A.M. de; Vliegen, Roy F.A.; Beets, Geerard L.; Papanikolaou, Nikos; Beets-Tan, Regina G.H.

    2016-01-01

    Purpose: Diffusion-weighted imaging (DWI) tumor volumetry is promising for rectal cancer response assessment, but an important drawback is that manual per-slice tumor delineation can be highly time consuming. This study investigated whether manual DWI-volumetry can be reproduced using a (semi)automated segmentation approach. Methods and Materials: Seventy-nine patients underwent magnetic resonance imaging (MRI) that included DWI (highest b value [b1000 or b1100]) before and after chemoradiation therapy (CRT). Tumor volumes were assessed on b1000 (or b1100) DWI before and after CRT by means of (1) automated segmentation (by 2 inexperienced readers), (2) semiautomated segmentation (manual adjustment of the volumes obtained by method 1 by 2 radiologists), and (3) manual segmentation (by 2 radiologists); this last assessment served as the reference standard. Intraclass correlation coefficients (ICC) and Dice similarity indices (DSI) were calculated to evaluate agreement between different methods and observers. Measurement times (from a radiologist's perspective) were recorded for each method. Results: Tumor volumes were not significantly different among the 3 methods, either before or after CRT (P=.08 to .92). ICCs compared to manual segmentation were 0.80 to 0.91 and 0.53 to 0.66 before and after CRT, respectively, for the automated segmentation and 0.91 to 0.97 and 0.61 to 0.75, respectively, for the semiautomated method. Interobserver agreement (ICC) pre and post CRT was 0.82 and 0.59 for automated segmentation, 0.91 and 0.73 for semiautomated segmentation, and 0.91 and 0.75 for manual segmentation, respectively. Mean DSI between the automated and semiautomated method were 0.83 and 0.58 pre-CRT and post-CRT, respectively; DSI between the automated and manual segmentation were 0.68 and 0.42 and 0.70 and 0.41 between the semiautomated and manual segmentation, respectively. Median measurement time for the radiologists was 0 seconds (pre- and post-CRT) for the

  6. Automated and Semiautomated Segmentation of Rectal Tumor Volumes on Diffusion-Weighted MRI: Can It Replace Manual Volumetry?

    Energy Technology Data Exchange (ETDEWEB)

    Heeswijk, Miriam M. van [Department of Radiology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Surgery, Maastricht University Medical Centre, Maastricht (Netherlands); Lambregts, Doenja M.J., E-mail: d.lambregts@nki.nl [Department of Radiology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Radiology, The Netherlands Cancer Institute, Amsterdam (Netherlands); Griethuysen, Joost J.M. van [GROW School for Oncology and Developmental Biology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Radiology, The Netherlands Cancer Institute, Amsterdam (Netherlands); Oei, Stanley [Department of Radiology, Maastricht University Medical Centre, Maastricht (Netherlands); Rao, Sheng-Xiang [Department of Radiology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Radiology, Zhongshan Hospital, Fudan University, Shanghai (China); Graaff, Carla A.M. de [Department of Radiology, Maastricht University Medical Centre, Maastricht (Netherlands); Vliegen, Roy F.A. [Atrium Medical Centre Parkstad/Zuyderland Medical Centre, Heerlen (Netherlands); Beets, Geerard L. [GROW School for Oncology and Developmental Biology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Surgery, The Netherlands Cancer Institute, Amsterdam (Netherlands); Papanikolaou, Nikos [Laboratory of Computational Medicine, Institute of Computer Science, FORTH, Heraklion, Crete (Greece); Beets-Tan, Regina G.H. [GROW School for Oncology and Developmental Biology, Maastricht University Medical Centre, Maastricht (Netherlands); Department of Radiology, The Netherlands Cancer Institute, Amsterdam (Netherlands)

    2016-03-15

    Purpose: Diffusion-weighted imaging (DWI) tumor volumetry is promising for rectal cancer response assessment, but an important drawback is that manual per-slice tumor delineation can be highly time consuming. This study investigated whether manual DWI-volumetry can be reproduced using a (semi)automated segmentation approach. Methods and Materials: Seventy-nine patients underwent magnetic resonance imaging (MRI) that included DWI (highest b value [b1000 or b1100]) before and after chemoradiation therapy (CRT). Tumor volumes were assessed on b1000 (or b1100) DWI before and after CRT by means of (1) automated segmentation (by 2 inexperienced readers), (2) semiautomated segmentation (manual adjustment of the volumes obtained by method 1 by 2 radiologists), and (3) manual segmentation (by 2 radiologists); this last assessment served as the reference standard. Intraclass correlation coefficients (ICC) and Dice similarity indices (DSI) were calculated to evaluate agreement between different methods and observers. Measurement times (from a radiologist's perspective) were recorded for each method. Results: Tumor volumes were not significantly different among the 3 methods, either before or after CRT (P=.08 to .92). ICCs compared to manual segmentation were 0.80 to 0.91 and 0.53 to 0.66 before and after CRT, respectively, for the automated segmentation and 0.91 to 0.97 and 0.61 to 0.75, respectively, for the semiautomated method. Interobserver agreement (ICC) pre and post CRT was 0.82 and 0.59 for automated segmentation, 0.91 and 0.73 for semiautomated segmentation, and 0.91 and 0.75 for manual segmentation, respectively. Mean DSI between the automated and semiautomated method were 0.83 and 0.58 pre-CRT and post-CRT, respectively; DSI between the automated and manual segmentation were 0.68 and 0.42 and 0.70 and 0.41 between the semiautomated and manual segmentation, respectively. Median measurement time for the radiologists was 0 seconds (pre- and post-CRT) for the

  7. Automated and Semiautomated Segmentation of Rectal Tumor Volumes on Diffusion-Weighted MRI: Can It Replace Manual Volumetry?

    Science.gov (United States)

    van Heeswijk, Miriam M; Lambregts, Doenja M J; van Griethuysen, Joost J M; Oei, Stanley; Rao, Sheng-Xiang; de Graaff, Carla A M; Vliegen, Roy F A; Beets, Geerard L; Papanikolaou, Nikos; Beets-Tan, Regina G H

    2016-03-15

    Diffusion-weighted imaging (DWI) tumor volumetry is promising for rectal cancer response assessment, but an important drawback is that manual per-slice tumor delineation can be highly time consuming. This study investigated whether manual DWI-volumetry can be reproduced using a (semi)automated segmentation approach. Seventy-nine patients underwent magnetic resonance imaging (MRI) that included DWI (highest b value [b1000 or b1100]) before and after chemoradiation therapy (CRT). Tumor volumes were assessed on b1000 (or b1100) DWI before and after CRT by means of (1) automated segmentation (by 2 inexperienced readers), (2) semiautomated segmentation (manual adjustment of the volumes obtained by method 1 by 2 radiologists), and (3) manual segmentation (by 2 radiologists); this last assessment served as the reference standard. Intraclass correlation coefficients (ICC) and Dice similarity indices (DSI) were calculated to evaluate agreement between different methods and observers. Measurement times (from a radiologist's perspective) were recorded for each method. Tumor volumes were not significantly different among the 3 methods, either before or after CRT (P=.08 to .92). ICCs compared to manual segmentation were 0.80 to 0.91 and 0.53 to 0.66 before and after CRT, respectively, for the automated segmentation and 0.91 to 0.97 and 0.61 to 0.75, respectively, for the semiautomated method. Interobserver agreement (ICC) pre and post CRT was 0.82 and 0.59 for automated segmentation, 0.91 and 0.73 for semiautomated segmentation, and 0.91 and 0.75 for manual segmentation, respectively. Mean DSI between the automated and semiautomated method were 0.83 and 0.58 pre-CRT and post-CRT, respectively; DSI between the automated and manual segmentation were 0.68 and 0.42 and 0.70 and 0.41 between the semiautomated and manual segmentation, respectively. Median measurement time for the radiologists was 0 seconds (pre- and post-CRT) for the automated method, 41 to 69 seconds (pre-CRT) and

  8. Program desk manual for occupational safety and health -- U.S. Department of Energy Richland Operations, Office of Environment Safety and Health

    International Nuclear Information System (INIS)

    Musen, L.G.

    1998-01-01

    The format of this manual is designed to make this valuable information easily accessible to the user as well as enjoyable to read. Each chapter contains common information such as Purpose, Scope, Policy and References, as well as information unique to the topic at hand. This manual can also be provided on a CD or Hanford Internet. Major topics include: Organization and program for operational safety; Occupational medicine; Construction and demolition; Material handling and storage; Hoisting and rigging; Explosives; Chemical hazards; Gas cylinders; Electrical; Boiler and pressure vessels; Industrial fire protection; Industrial hygiene; and Safety inspection checklist

  9. Program desk manual for occupational safety and health -- U.S. Department of Energy Richland Operations, Office of Environment Safety and Health

    Energy Technology Data Exchange (ETDEWEB)

    Musen, L.G.

    1998-08-27

    The format of this manual is designed to make this valuable information easily accessible to the user as well as enjoyable to read. Each chapter contains common information such as Purpose, Scope, Policy and References, as well as information unique to the topic at hand. This manual can also be provided on a CD or Hanford Internet. Major topics include: Organization and program for operational safety; Occupational medicine; Construction and demolition; Material handling and storage; Hoisting and rigging; Explosives; Chemical hazards; Gas cylinders; Electrical; Boiler and pressure vessels; Industrial fire protection; Industrial hygiene; and Safety inspection checklist.

  10. Sequence Coding and Search System for licensee event reports: coder's manual. Volume 4

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This four volume report documents and describes SCSS in detail. Volume 3 and 4 provide a technical processor, new to SCSS, the information and methodology necessary to capture descriptive data from the LER and to codify that data into a structured format and serve as reference material for the more experienced technical processor, and contains information that is essential for the more advanced user who needs to be familiar with the intricate coding techniques in order to retrieve specific details in a sequence. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 4

  11. Sequence Coding and Search System for licensee event reports: coder's manual. Volume 3

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This four volume report documents and describes SCSS in detail. Volumes 3 and 4 provide a technical processor, new to SCSS, the information and methodology necessary to capture descriptive data from the LER and to codify that data into a structured format and serve as reference material for the more experienced technical processor, and contains information is essential for the more advanced user who needs to be familiar with the intricate coding techniques in order to retrieve specific details in a sequence. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 3

  12. Chemical Safety Vulnerability Working Group report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms.

  13. Chemical Safety Vulnerability Working Group report. Volume 1

    International Nuclear Information System (INIS)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms

  14. Chemical Safety Vulnerability Working Group report. Volume 3

    International Nuclear Information System (INIS)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 3 consists of eleven appendices containing the following: Field verification reports for Idaho National Engineering Lab., Rocky Flats Plant, Brookhaven National Lab., Los Alamos National Lab., and Sandia National Laboratories (NM); Mini-visits to small DOE sites; Working Group meeting, June 7--8, 1994; Commendable practices; Related chemical safety initiatives at DOE; Regulatory framework and industry initiatives related to chemical safety; and Chemical inventory data from field self-evaluation reports

  15. Chemical Safety Vulnerability Working Group report. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 3 consists of eleven appendices containing the following: Field verification reports for Idaho National Engineering Lab., Rocky Flats Plant, Brookhaven National Lab., Los Alamos National Lab., and Sandia National Laboratories (NM); Mini-visits to small DOE sites; Working Group meeting, June 7--8, 1994; Commendable practices; Related chemical safety initiatives at DOE; Regulatory framework and industry initiatives related to chemical safety; and Chemical inventory data from field self-evaluation reports.

  16. Age-related differences in regional brain volumes: A comparison of optimized voxel-based morphometry to manual volumetry

    Science.gov (United States)

    Kennedy, Kristen M.; Erickson, Kirk I.; Rodrigue, Karen M.; Voss, Michelle W.; Colcombe, Stan J.; Kramer, Arthur F.; Acker, James D.; Raz, Naftali

    2009-01-01

    Regional manual volumetry is the gold standard of in vivo neuroanatomy, but is labor-intensive, can be imperfectly reliable, and allows for measuring limited number of regions. Voxel-based morphometry (VBM) has perfect repeatability and assesses local structure across the whole brain. However, its anatomic validity is unclear, and with its increasing popularity, a systematic comparison of VBM to manual volumetry is necessary. The few existing comparison studies are limited by small samples, qualitative comparisons, and limited selection and modest reliability of manual measures. Our goal was to overcome those limitations by quantitatively comparing optimized VBM findings with highly reliable multiple regional measures in a large sample (N = 200) across a wide agespan (18–81). We report a complex pattern of similarities and differences. Peak values of VBM volume estimates (modulated density) produced stronger age differences and a different spatial distribution from manual measures. However, when we aggregated VBM-derived information across voxels contained in specific anatomically defined regions (masks), the patterns of age differences became more similar, although important discrepancies emerged. Notably, VBM revealed stronger age differences in the regions bordering CSF and white matter areas prone to leukoaraiosis, and VBM was more likely to report nonlinearities in age-volume relationships. In the white matter regions, manual measures showed stronger negative associations with age than the corresponding VBM-based masks. We conclude that VBM provides realistic estimates of age differences in the regional gray matter only when applied to anatomically defined regions, but overestimates effects when individual peaks are interpreted. It may be beneficial to use VBM as a first-pass strategy, followed by manual measurement of anatomically-defined regions. PMID:18276037

  17. Radiological Safety Analysis Computer (RSAC) Program Version 7.0 Users’ Manual

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Bradley J Schrader

    2009-03-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.0 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users’ manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods.

  18. Radiological Safety Analysis Computer (RSAC) Program Version 7.2 Users’ Manual

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Bradley J Schrader

    2010-10-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.2 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users’ manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods.

  19. Radiological Safety Analysis Computer (RSAC) Program Version 7.0 Users Manual

    International Nuclear Information System (INIS)

    Schrader, Bradley J.

    2009-01-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.0 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods

  20. Federal Radiological Monitoring and Assessment Center Monitoring Manual Volume 1, Operations

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Aerial Measurement Systems

    2012-07-31

    The Monitoring division is primarily responsible for the coordination and direction of: Aerial measurements to delineate the footprint of radioactive contaminants that have been released into the environment. Monitoring of radiation levels in the environment; Sampling to determine the extent of contaminant deposition in soil, water, air and on vegetation; Preliminary field analyses to quantify soil concentrations or depositions; and Environmental and personal dosimetry for FRMAC field personnel, during a Consequence Management Response Team (CMRT) and Federal Radiological Monitoring and Assessment Center (FRMAC) response. Monitoring and sampling techniques used during CM/FRMAC operations are specifically selected for use during radiological emergencies where large numbers of measurements and samples must be acquired, analyzed, and interpreted in the shortest amount of time possible. In addition, techniques and procedures are flexible so that they can be used during a variety of different scenarios; e.g., accidents involving releases from nuclear reactors, contamination by nuclear waste, nuclear weapon accidents, space vehicle reentries, or contamination from a radiological dispersal device. The Monitoring division also provides technicians to support specific Health and Safety Division activities including: The operation of the Hotline; FRMAC facility surveys; Assistance with Health and Safety at Check Points; and Assistance at population assembly areas which require support from the FRMAC. This volume covers deployment activities, initial FRMAC activities, development and implementation of the monitoring and assessment plan, the briefing of field teams, and the transfer of FRMAC to the EPA.

  1. The Container Tree Nursery Manual: Volume 7, Seedling processing, storage, and outplanting

    Science.gov (United States)

    Thomas D. Landis; R. Kasten Dumroese; Diane L. Haase

    2010-01-01

    This manual is based on the best current knowledge of container nursery management and should be used as a general reference. Recommendations were made using the best information available at the time and are, therefore, subject to revision as more knowledge becomes available. Much of the information in this manual was primarily developed from information on growing...

  2. SCDAP/RELAP5/MOD 3.1 code manual: User's guide and input manual. Volume 3

    International Nuclear Information System (INIS)

    Coryell, E.W.; Johnsen, E.C.; Allison, C.M.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume provides guidelines to code users based upon lessons learned during the developmental assessment process. A description of problem control and the installation process is included. Appendix a contains the description of the input requirements

  3. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0. Volume 5, Systems Analysis and Risk Assessment (SARA) tutorial manual

    International Nuclear Information System (INIS)

    Sattison, M.B.; Russell, K.D.; Skinner, N.L.

    1994-07-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs) primarily for nuclear power plants. This volume is the tutorial manual for the Systems Analysis and Risk Assessment (SARA) System Version 5.0, a microcomputer-based system used to analyze the safety issues of a open-quotes familyclose quotes [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons is provided that guides the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis capabilities

  4. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations

    International Nuclear Information System (INIS)

    Lacour, J.

    1964-01-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [fr

  5. High School and Beyond. 1980 Senior Coort. Third-Follow-Up (1986). Data File User's Manual. Volume II: Survey Instruments. Contractor Report.

    Science.gov (United States)

    Sebring, Penny; And Others

    Survey instruments used in the collection of data for the High School and Beyond base year (1980) through the third follow-up surveys (1986) are provided as Volume II of a user's manual for the senior cohort data file. The complete user's manual is designed to provide the extensive documentation necessary for using the cohort data files. Copies of…

  6. Technical reference manual for TIME4. Version 1.0, Volume 1

    International Nuclear Information System (INIS)

    Wilmot, R.D.; Ringrose, P.S.; Larkin, J.P.A.; Kleissen, F.A.T.

    1991-07-01

    This document is the Technical Reference Manual for the TIME4 model. TIME4 is the environmental change model developed for use in the probabilistic risk analysis of deep disposal of radioactive waste. The Technical Reference Manual describes the theoretical background to the model. The modelling method is described, followed by a review of related work and a detailed description for each sub-model. (author)

  7. GENII (Generation II): The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

    Energy Technology Data Exchange (ETDEWEB)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-09-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs.

  8. GENII [Generation II]: The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

    International Nuclear Information System (INIS)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-09-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs

  9. Associations Between Family Adversity and Brain Volume in Adolescence: Manual vs. Automated Brain Segmentation Yields Different Results

    Directory of Open Access Journals (Sweden)

    Hannah Lyden

    2016-09-01

    Full Text Available Associations between brain structure and early adversity have been inconsistent in the literature. These inconsistencies may be partially due to methodological differences. Different methods of brain segmentation may produce different results, obscuring the relationship between early adversity and brain volume. Moreover, adolescence is a time of significant brain growth and certain brain areas have distinct rates of development, which may compromise the accuracy of automated segmentation approaches. In the current study, 23 adolescents participated in two waves of a longitudinal study. Family aggression was measured when the youths were 12 years old, and structural scans were acquired an average of 4 years later. Bilateral amygdalae and hippocampi were segmented using three different methods (manual tracing, FSL, and NeuroQuant. The segmentation estimates were compared, and linear regressions were run to assess the relationship between early family aggression exposure and all three volume segmentation estimates. Manual tracing results showed a positive relationship between family aggression and right amygdala volume, whereas FSL segmentation showed negative relationships between family aggression and both the left and right hippocampi. However, results indicate poor overlap between methods, and different associations between early family aggression exposure and brain volume depending on the segmentation method used.

  10. Associations between Family Adversity and Brain Volume in Adolescence: Manual vs. Automated Brain Segmentation Yields Different Results.

    Science.gov (United States)

    Lyden, Hannah; Gimbel, Sarah I; Del Piero, Larissa; Tsai, A Bryna; Sachs, Matthew E; Kaplan, Jonas T; Margolin, Gayla; Saxbe, Darby

    2016-01-01

    Associations between brain structure and early adversity have been inconsistent in the literature. These inconsistencies may be partially due to methodological differences. Different methods of brain segmentation may produce different results, obscuring the relationship between early adversity and brain volume. Moreover, adolescence is a time of significant brain growth and certain brain areas have distinct rates of development, which may compromise the accuracy of automated segmentation approaches. In the current study, 23 adolescents participated in two waves of a longitudinal study. Family aggression was measured when the youths were 12 years old, and structural scans were acquired an average of 4 years later. Bilateral amygdalae and hippocampi were segmented using three different methods (manual tracing, FSL, and NeuroQuant). The segmentation estimates were compared, and linear regressions were run to assess the relationship between early family aggression exposure and all three volume segmentation estimates. Manual tracing results showed a positive relationship between family aggression and right amygdala volume, whereas FSL segmentation showed negative relationships between family aggression and both the left and right hippocampi. However, results indicate poor overlap between methods, and different associations were found between early family aggression exposure and brain volume depending on the segmentation method used.

  11. GENII: The Hanford Environmental Radiation Dosimetry Software System: Volume 2, Users' manual: Hanford Environmental Dosimetry Upgrade Project

    International Nuclear Information System (INIS)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-11-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). The purpose of this coupled system of computer codes is to analyze environmental contamination of, air, water, or soil. This is accomplished by calculating radiation doses to individuals or populations. GENII is described in three volumes of documentation. This second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. The first volume describes the theoretical considerations of the system. The third volume is a Code Maintenance Manual for the user who requires knowledge of code detail. It includes logic diagrams, global dictionary, worksheets, example hand calculations, and listings of the code and its associated data libraries. 27 refs., 17 figs., 23 tabs

  12. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  13. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  14. Chemical Safety Vulnerability Working Group report. Volume 2

    International Nuclear Information System (INIS)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site

  15. Chemical Safety Vulnerability Working Group report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site.

  16. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  17. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  18. Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 2, Investigators`s Manual

    Energy Technology Data Exchange (ETDEWEB)

    Paradies, M.; Unger, L. [System Improvements, Inc., Knoxville, TN (United States); Haas, P.; Terranova, M. [Concord Associates, Inc., Knoxville, TN (United States)

    1993-10-01

    The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause event at nuclear power plants. This document, Volume II, is a field manual for use by investigators when performing event investigations. Volume II includes the HPIP Procedure, the HPIP Modules, and Appendices that provide extensive documentation of each investigation technique.

  19. Broadband Fan Noise Prediction System for Turbofan Engines. Volume 1; Setup_BFaNS User's Manual and Developer's Guide

    Science.gov (United States)

    Morin, Bruce L.

    2010-01-01

    Pratt & Whitney has developed a Broadband Fan Noise Prediction System (BFaNS) for turbofan engines. This system computes the noise generated by turbulence impinging on the leading edges of the fan and fan exit guide vane, and noise generated by boundary-layer turbulence passing over the fan trailing edge. BFaNS has been validated on three fan rigs that were tested during the NASA Advanced Subsonic Technology Program (AST). The predicted noise spectra agreed well with measured data. The predicted effects of fan speed, vane count, and vane sweep also agreed well with measurements. The noise prediction system consists of two computer programs: Setup_BFaNS and BFaNS. Setup_BFaNS converts user-specified geometry and flow-field information into a BFaNS input file. From this input file, BFaNS computes the inlet and aft broadband sound power spectra generated by the fan and FEGV. The output file from BFaNS contains the inlet, aft and total sound power spectra from each noise source. This report is the first volume of a three-volume set documenting the Broadband Fan Noise Prediction System: Volume 1: Setup_BFaNS User s Manual and Developer s Guide; Volume 2: BFaNS User's Manual and Developer s Guide; and Volume 3: Validation and Test Cases. The present volume begins with an overview of the Broadband Fan Noise Prediction System, followed by step-by-step instructions for installing and running Setup_BFaNS. It concludes with technical documentation of the Setup_BFaNS computer program.

  20. Broadband Fan Noise Prediction System for Turbofan Engines. Volume 2; BFaNS User's Manual and Developer's Guide

    Science.gov (United States)

    Morin, Bruce L.

    2010-01-01

    Pratt & Whitney has developed a Broadband Fan Noise Prediction System (BFaNS) for turbofan engines. This system computes the noise generated by turbulence impinging on the leading edges of the fan and fan exit guide vane, and noise generated by boundary-layer turbulence passing over the fan trailing edge. BFaNS has been validated on three fan rigs that were tested during the NASA Advanced Subsonic Technology Program (AST). The predicted noise spectra agreed well with measured data. The predicted effects of fan speed, vane count, and vane sweep also agreed well with measurements. The noise prediction system consists of two computer programs: Setup_BFaNS and BFaNS. Setup_BFaNS converts user-specified geometry and flow-field information into a BFaNS input file. From this input file, BFaNS computes the inlet and aft broadband sound power spectra generated by the fan and FEGV. The output file from BFaNS contains the inlet, aft and total sound power spectra from each noise source. This report is the second volume of a three-volume set documenting the Broadband Fan Noise Prediction System: Volume 1: Setup_BFaNS User s Manual and Developer s Guide; Volume 2: BFaNS User s Manual and Developer s Guide; and Volume 3: Validation and Test Cases. The present volume begins with an overview of the Broadband Fan Noise Prediction System, followed by step-by-step instructions for installing and running BFaNS. It concludes with technical documentation of the BFaNS computer program.

  1. Hanford Environmental Information System (HEIS) Operator`s Manual. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Schreck, R.I.

    1991-10-01

    The Hanford Environmental Information System (HEIS) is a consolidated set of automated resources that effectively manage the data gathered during environmental monitoring and restoration of the Hanford Site. The HEIS includes an integrated database that provides consistent and current data to all users and promotes sharing of data by the entire user community. This manual describes the facilities available to the operational user who is responsible for data entry, processing, scheduling, reporting, and quality assurance. A companion manual, the HEIS User`s Manual, describes the facilities available-to the scientist, engineer, or manager who uses the system for environmental monitoring, assessment, and restoration planning; and to the regulator who is responsible for reviewing Hanford Site operations against regulatory requirements and guidelines.

  2. Fuel Rod Consolidation Project: Phase 2, Final report: Volume 5, Operations and maintenance manual

    International Nuclear Information System (INIS)

    1988-01-01

    The purpose of this manual is to describe the function, installation, operation and maintenance of the Fuel Rod Consolidation System. This Document is preliminary and must be updated to incorporate any modifications to the mechanical and electrical systems that are performed during construction. Any changes and specific references related to the software requirements will be provided as the software is developed in Phase III. Setpoints related to equipment positions as a function of resolver and position transducer readings will also be provided in Phase III. References such as vendor supplied Operating and Maintenance Manuals for vendor components and assemblies are not available until a receipt of a purchase order. These references will become an integral part of this manual during the construction phase

  3. Validation of hippocampal volumes measured using a manual method and two automated methods (FreeSurfer and IBASPM) in chronic major depressive disorder

    Energy Technology Data Exchange (ETDEWEB)

    Tae, Woo Suk; Lee, Kang Uk; Nam, Eui-Cheol; Kim, Keun Woo [Kangwon National University College of Medicine, Neuroscience Research Institute, Kangwon (Korea); Kim, Sam Soo [Kangwon National University College of Medicine, Neuroscience Research Institute, Kangwon (Korea); Kangwon National University Hospital, Department of Radiology, Kangwon-do (Korea)

    2008-07-15

    To validate the usefulness of the packages available for automated hippocampal volumetry, we measured hippocampal volumes using one manual and two recently developed automated volumetric methods. The study included T1-weighted magnetic resonance imaging (MRI) of 21 patients with chronic major depressive disorder (MDD) and 20 normal controls. Using coronal turbo field echo (TFE) MRI with a slice thickness of 1.3 mm, the hippocampal volumes were measured using three methods: manual volumetry, surface-based parcellation using FreeSurfer, and individual atlas-based volumetry using IBASPM. In addition, the intracranial cavity volume (ICV) was measured manually. The absolute left hippocampal volume of the patients with MDD measured using all three methods was significantly smaller than the left hippocampal volume of the normal controls (manual P=0.029, FreeSurfer P=0.035, IBASPM P=0.018). After controlling for the ICV, except for the right hippocampal volume measured using FreeSurfer, both measured hippocampal volumes of the patients with MDD were significantly smaller than the measured hippocampal volumes of the normal controls (right manual P=0.019, IBASPM P=0.012; left manual P=0.003, FreeSurfer P=0.010, IBASPM P=0.002). In the intrarater reliability test, the intraclass correlation coefficients (ICCs) were all excellent (manual right 0.947, left 0.934; FreeSurfer right 1.000, left 1.000; IBASPM right 1.000, left 1.000). In the test of agreement between the volumetric methods, the ICCs were right 0.846 and left 0.848 (manual and FreeSurfer), and right 0.654 and left 0.717 (manual and IBASPM). The automated hippocampal volumetric methods showed good agreement with manual hippocampal volumetry, but the volume measured using FreeSurfer was 35% larger and the agreement was questionable with IBASPM. Although the automated methods could detect hippocampal atrophy in the patients with MDD, the results indicate that manual hippocampal volumetry is still the gold standard

  4. Validation of hippocampal volumes measured using a manual method and two automated methods (FreeSurfer and IBASPM) in chronic major depressive disorder

    International Nuclear Information System (INIS)

    Tae, Woo Suk; Lee, Kang Uk; Nam, Eui-Cheol; Kim, Keun Woo; Kim, Sam Soo

    2008-01-01

    To validate the usefulness of the packages available for automated hippocampal volumetry, we measured hippocampal volumes using one manual and two recently developed automated volumetric methods. The study included T1-weighted magnetic resonance imaging (MRI) of 21 patients with chronic major depressive disorder (MDD) and 20 normal controls. Using coronal turbo field echo (TFE) MRI with a slice thickness of 1.3 mm, the hippocampal volumes were measured using three methods: manual volumetry, surface-based parcellation using FreeSurfer, and individual atlas-based volumetry using IBASPM. In addition, the intracranial cavity volume (ICV) was measured manually. The absolute left hippocampal volume of the patients with MDD measured using all three methods was significantly smaller than the left hippocampal volume of the normal controls (manual P=0.029, FreeSurfer P=0.035, IBASPM P=0.018). After controlling for the ICV, except for the right hippocampal volume measured using FreeSurfer, both measured hippocampal volumes of the patients with MDD were significantly smaller than the measured hippocampal volumes of the normal controls (right manual P=0.019, IBASPM P=0.012; left manual P=0.003, FreeSurfer P=0.010, IBASPM P=0.002). In the intrarater reliability test, the intraclass correlation coefficients (ICCs) were all excellent (manual right 0.947, left 0.934; FreeSurfer right 1.000, left 1.000; IBASPM right 1.000, left 1.000). In the test of agreement between the volumetric methods, the ICCs were right 0.846 and left 0.848 (manual and FreeSurfer), and right 0.654 and left 0.717 (manual and IBASPM). The automated hippocampal volumetric methods showed good agreement with manual hippocampal volumetry, but the volume measured using FreeSurfer was 35% larger and the agreement was questionable with IBASPM. Although the automated methods could detect hippocampal atrophy in the patients with MDD, the results indicate that manual hippocampal volumetry is still the gold standard

  5. The procedures manual of the Environmental Measurements Laboratory. Volume 1, 28. edition

    Energy Technology Data Exchange (ETDEWEB)

    Chieco, N.A. [ed.

    1997-02-01

    This manual covers procedures and technology currently in use at the Environmental Measurements Laboratory. An attempt is made to be sure that all work carried out will be of the highest quality. Attention is focused on the following areas: quality assurance; sampling; radiation measurements; analytical chemistry; radionuclide data; special facilities; and specifications.

  6. The procedures manual of the Environmental Measurements Laboratory. Volume 1, 28. edition

    International Nuclear Information System (INIS)

    Chieco, N.A.

    1997-02-01

    This manual covers procedures and technology currently in use at the Environmental Measurements Laboratory. An attempt is made to be sure that all work carried out will be of the highest quality. Attention is focused on the following areas: quality assurance; sampling; radiation measurements; analytical chemistry; radionuclide data; special facilities; and specifications

  7. Manual on radiation protection in hospital and general practice. Volume 4. Radiation protection in dentistry

    Energy Technology Data Exchange (ETDEWEB)

    Koren, K; Wuehrmann, A H

    1977-01-01

    The nine chapters of this manual on radiation protection in dentistry discuss the following topics: the need for radiation protection; delegation of responsibility; radiographic equipment; radiographic film; radiographic techniques; film processing and handling; patient doses; general radiation protection and monitoring; and educational standards. (HLW)

  8. Viscous wing theory development. Volume 2: GRUMWING computer program user's manual

    Science.gov (United States)

    Chow, R. R.; Ogilvie, P. L.

    1986-01-01

    This report is a user's manual which describes the operation of the computer program, GRUMWING. The program computes the viscous transonic flow over three-dimensional wings using a boundary layer type viscid-inviscid interaction approach. The inviscid solution is obtained by an approximate factorization (AFZ)method for the full potential equation. The boundary layer solution is based on integral entrainment methods.

  9. Texas trip generation manual : 1st edition-volume 1 : user's guide.

    Science.gov (United States)

    2014-08-01

    The purpose of this Manual is to provide a summary of Texas trip generation data for various : Land Use Codes (LUCs) and time periods, for data obtained from workplace and special : generator (WSG) surveys performed as part of the Texas Travel Survey...

  10. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  11. Flight dynamics analysis and simulation of heavy lift airships. Volume 2: Technical manual

    Science.gov (United States)

    Ringland, R. F.; Tischler, M. B.; Jex, H. R.; Emmen, R. D.; Ashkenas, I. L.

    1982-01-01

    The mathematical models embodied in the simulation are described in considerable detail and with supporting evidence for the model forms chosen. In addition the trimming and linearization algorithms used in the simulation are described. Appendices to the manual identify reference material for estimating the needed coefficients for the input data and provide example simulation results.

  12. School Survey on Crime and Safety (SSOCS) 2000 Public-Use Data Files, User's Manual, and Detailed Data Documentation. [CD-ROM].

    Science.gov (United States)

    National Center for Education Statistics (ED), Washington, DC.

    This CD-ROM contains the raw, public-use data from the 2000 School Survey on Crime and Safety (SSOCS) along with a User's Manual and Detailed Data Documentation. The data are provided in SAS, SPSS, STATA, and ASCII formats. The User's Manual and the Detailed Data Documentation are provided as .pdf files. (Author)

  13. A technique for manual definition of an irregular volume of interest in single photon emission computed tomography

    International Nuclear Information System (INIS)

    Fleming, J.S.; Kemp, P.M.; Bolt, L.

    1999-01-01

    A technique is described for manually outlining a volume of interest (VOI) in a three-dimensional SPECT dataset. Regions of interest (ROIs) are drawn on three orthogonal maximum intensity projections. Image masks based on these ROIs are backprojected through the image volume and the resultant 3D dataset is segmented to produce the VOI. The technique has been successfully applied in the exclusion of unwanted areas of activity adjacent to the brain when segmenting the organ in SPECT imaging using 99m Tc HMPAO. An example of its use for segmentation in tumour imaging is also presented. The technique is of value for applications involving semi-automatic VOI definition in SPECT. (author)

  14. Instrumentation and control upgrade evaluation methodology. Volume 1: Manual. Final report

    International Nuclear Information System (INIS)

    Bliss, M.; Brown, E.; Florio, F.; Stofko, M.

    1996-07-01

    This methodology manual describes how to develop an Upgrade Evaluation Report (UER) for an I and C system that has been identified as an upgrade candidate in the I and C Life Cycle Management Plan (LCMP). A UER can be developed by a nuclear power plant to determine the cost-effectiveness and feasibility of upgrading an aging or obsolete Instrumentation and Control (I and C) System. A separate UER is developed for each upgrade candidate system. A UER will determine if a given system requires an upgrade or if it is more cost-effective to maintain the present system. If an upgrade is unnecessary, the system will be reclassified as a retained system and a system maintenance plan will be developed for it. This manual is accompanied by a workbook (EPRI TR-104963-V2) which contains various worksheets, outlines, and generic interview questions that aid in the UER development process

  15. Instrumentation and Control Life Cycle Management Plan Methodology. Volume 1, Manual: Final report

    International Nuclear Information System (INIS)

    Quick, D.S.; Murray, S.; Florio, F.; Bliss, M.J.

    1995-08-01

    This methodology manual describes how to develop a Life Cycle Management Plan (LCMP). An LCMP is a long-term strategic plan that can be developed for a nuclear power plant to cost-effectively maintain and upgrade its aging or obsolete Instrumentation and Control (I ampersand C) systems. An LCMP defines the utility's mission and objectives in regards to long range I ampersand C planning, as well as the plant's present configuration (I ampersand C systems, networks, man machine interfaces, etc.), its desired future I ampersand C systems, a long term I ampersand C maintenance strategy, and initial upgrade priorities and schedules to cost-effectively implement system upgrades. This manual is accompanied by a workbook (EPRI TR-105555-V2) which contains various worksheets, outlines, and generic interview questions that aid in the LCNW development process

  16. Main-coolant-pump shaft-seal guidelines. Volume 1. Maintenance-manual guidelines. Final report

    International Nuclear Information System (INIS)

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and a listing of information and data which should be included in maintenance manuals and procedures for Main Coolant Pump Shaft Seals. The noted guidelines and data listing are developed from EPRI sponsored nuclear plant seal operating experience studies. The maintenance oriented results of the most recent such study is summarized. The shaft seal and its auxiliary supporting systems are discussed from both technical and maintenance related viewpoints

  17. Documentation for Grants Equal To Tax model: Volume 2, User manual

    International Nuclear Information System (INIS)

    1986-01-01

    The Grants Equal To Tax (GETT) model was developed to assist in evaluating the amount of Federal grant monies that would go to State and local jurisdictions as a result of site characterization and repository activities. Because a variety of policy issues are involved in estimating the grant payments, the GETT model is designed to have the flexibility to implement different policy options. The GETT User Manual describes how to run the GETT model and how to modify the data bases. 22 figs

  18. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2

    International Nuclear Information System (INIS)

    Osborn, R.N.; Olson, J.; Sommers, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators

  19. RELAP5/MOD3 code manual. Volume 4, Models and correlations

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations

  20. Tank Farms Technical Safety Requirements. Volume 1 and 2

    International Nuclear Information System (INIS)

    CASH, R.J.

    2000-01-01

    The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR)

  1. Controlled air incinerator for radioactive waste. Volume II. Engineering design references manual

    International Nuclear Information System (INIS)

    Koenig, R.A.; Draper, W.E.; Newmyer, J.M.; Warner, C.L.

    1982-11-01

    This two-volume report is a detailed design and operating documentation of the Los Alamos National Laboratory Controlled Air Incinerator (CAI) and is an aid to technology transfer to other Department of Energy contractor sites and the commercial sector. Volume I describes the CAI process, equipment, and performance, and it recommends modifications based on Los Alamos experience. It provides the necessary information for conceptual design and feasibility studies. Volume II provides descriptive engineering information such as drawings, specifications, calculations, and costs. It aids duplication of the process at other facilities

  2. Los Alamos Controlled Air Incinerator for radioactive waste. Volume II. Engineering design reference manual

    Energy Technology Data Exchange (ETDEWEB)

    Koenig, R.A.; Draper, W.E.; Newmyer, J.M.; Warner, C.L.

    1982-10-01

    This two-volume report is a detailed design and operating documentation of the Los Alamos National Laboratory Controlled Air Incinerator (CAI) and is an aid to technology transfer to other Department of Energy contractor sites and the commercial sector. Volume I describes the CAI process, equipment, and performance, and it recommends modifications based on Los Alamos experience. It provides the necessary information for conceptual design and feasibility studies. Volume II provides descriptive engineering information such as drawings, specifications, calculations, and costs. It aids duplication of the process at other facilities.

  3. Los Alamos Controlled Air Incinerator for radioactive waste. Volume II. Engineering design reference manual

    International Nuclear Information System (INIS)

    Koenig, R.A.; Draper, W.E.; Newmyer, J.M.; Warner, C.L.

    1982-10-01

    This two-volume report is a detailed design and operating documentation of the Los Alamos National Laboratory Controlled Air Incinerator (CAI) and is an aid to technology transfer to other Department of Energy contractor sites and the commercial sector. Volume I describes the CAI process, equipment, and performance, and it recommends modifications based on Los Alamos experience. It provides the necessary information for conceptual design and feasibility studies. Volume II provides descriptive engineering information such as drawings, specifications, calculations, and costs. It aids duplication of the process at other facilities

  4. PETRO-SAFE '94 conference papers: Book 2. Volume 5: Emergency response ampersand spill control; Volume 6: Remediation; Volume 7: Health ampersand safety issues

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    The Fifth Annual Environmental, Safety and Health Conference and Exhibition for the oil, gas and petrochemical industries was held January 25--27, 1994 in Houston, Texas. The objective of this conference was to provide a multidisciplinary forum dealing with state-of-the-art environmental and safety issues. This volume focuses on the following: emergency response and spill control; remediation; and health and safety issues. Individual papers have been processed separately for inclusion in the appropriate data bases

  5. BPACK -- A computer model package for boiler reburning/co-firing performance evaluations. User`s manual, Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Wu, K.T.; Li, B.; Payne, R.

    1992-06-01

    This manual presents and describes a package of computer models uniquely developed for boiler thermal performance and emissions evaluations by the Energy and Environmental Research Corporation. The model package permits boiler heat transfer, fuels combustion, and pollutant emissions predictions related to a number of practical boiler operations such as fuel-switching, fuels co-firing, and reburning NO{sub x} reductions. The models are adaptable to most boiler/combustor designs and can handle burner fuels in solid, liquid, gaseous, and slurried forms. The models are also capable of performing predictions for combustion applications involving gaseous-fuel reburning, and co-firing of solid/gas, liquid/gas, gas/gas, slurry/gas fuels. The model package is conveniently named as BPACK (Boiler Package) and consists of six computer codes, of which three of them are main computational codes and the other three are input codes. The three main codes are: (a) a two-dimensional furnace heat-transfer and combustion code: (b) a detailed chemical-kinetics code; and (c) a boiler convective passage code. This user`s manual presents the computer model package in two volumes. Volume 1 describes in detail a number of topics which are of general users` interest, including the physical and chemical basis of the models, a complete description of the model applicability, options, input/output, and the default inputs. Volume 2 contains a detailed record of the worked examples to assist users in applying the models, and to illustrate the versatility of the codes.

  6. GO software version 3.0: Volume 1, Overview: Computer Code Manual

    International Nuclear Information System (INIS)

    1988-06-01

    The GO Methodology is a success-oriented probabilistic system performance analysis technique. The methodology can be used to quantify system reliability and availability, identify and rank critical components and the contributors to system failure, construct event trees, perform statistical uncertainty analysis, and evaluate the effects of external events and common cause failures on system performance. This Overview Manual provides a description of the GO methodology, how it can be used, the benefits of using it in the analysis of complex systems, and a comparison of the methodology with fault tree analysis. 14 refs., 35 figs., 9 tabs

  7. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  8. Propensity scores-potential outcomes framework to incorporate severity probabilities in the highway safety manual crash prediction algorithm.

    Science.gov (United States)

    Sasidharan, Lekshmi; Donnell, Eric T

    2014-10-01

    Accurate estimation of the expected number of crashes at different severity levels for entities with and without countermeasures plays a vital role in selecting countermeasures in the framework of the safety management process. The current practice is to use the American Association of State Highway and Transportation Officials' Highway Safety Manual crash prediction algorithms, which combine safety performance functions and crash modification factors, to estimate the effects of safety countermeasures on different highway and street facility types. Many of these crash prediction algorithms are based solely on crash frequency, or assume that severity outcomes are unchanged when planning for, or implementing, safety countermeasures. Failing to account for the uncertainty associated with crash severity outcomes, and assuming crash severity distributions remain unchanged in safety performance evaluations, limits the utility of the Highway Safety Manual crash prediction algorithms in assessing the effect of safety countermeasures on crash severity. This study demonstrates the application of a propensity scores-potential outcomes framework to estimate the probability distribution for the occurrence of different crash severity levels by accounting for the uncertainties associated with them. The probability of fatal and severe injury crash occurrence at lighted and unlighted intersections is estimated in this paper using data from Minnesota. The results show that the expected probability of occurrence of fatal and severe injury crashes at a lighted intersection was 1 in 35 crashes and the estimated risk ratio indicates that the respective probabilities at an unlighted intersection was 1.14 times higher compared to lighted intersections. The results from the potential outcomes-propensity scores framework are compared to results obtained from traditional binary logit models, without application of propensity scores matching. Traditional binary logit analysis suggests that

  9. Using a profiling process to insure program quality: Volume I - a self-instructional manual

    Energy Technology Data Exchange (ETDEWEB)

    Kaser, J.S.; Roody, D.S.; Raizen, S.A.

    1996-11-01

    Between 1990 and 1995 the U.S. Department of Energy (DOE) and the National Center for Improving Science Education (National Center) developed a system for ongoing evaluation of K-12 educational programs in the DOE-supported national energy Laboratories. As part of the formative evaluation component of this collaborative endeavor, field staff in the Laboratories began creating profiles of their programs. However, many individuals within DOE Headquarters were not familiar with this profiling process and were unprepared to use the valuable information that the profiles generated. This manual was produced to orient Headquarters staff to profiling. It focuses on how Headquarters staff can use the profiling process to help their funded programs establish and/or maintain high quality. Its purpose, then, is not to train Headquarters staff to become proficient in profiling, but to show them how to draw on the Laboratories` use of profiling to bring about program improvement. Profiling is the process of systematically examining and describing a program`s elements against a set of components that define Effective Practice. The instrument used to capture the data for analysis is called a template, and most of this manual focuses on the templates and how to read and interpret them. However, since it is important to understand these data in context, the authors also describe what should accompany each template in a complete profiling packet and offer guidelines for reviewing complete packets and providing feedback to program managers.

  10. Evaluation of left atrial function by multidetector computed tomography before left atrial radiofrequency-catheter ablation: Comparison of a manual and automated 3D volume segmentation method

    International Nuclear Information System (INIS)

    Wolf, Florian; Ourednicek, Petr; Loewe, Christian; Richter, Bernhard; Goessinger, Heinz David; Gwechenberger, Marianne; Plank, Christina; Schernthaner, Ruediger Egbert; Toepker, Michael; Lammer, Johannes; Feuchtner, Gudrun M.

    2010-01-01

    Introduction: The purpose of this study was to compare a manual and automated 3D volume segmentation tool for evaluation of left atrial (LA) function by 64-slice multidetector-CT (MDCT). Methods and materials: In 33 patients with paroxysmal atrial fibrillation a MDCT scan was performed before radiofrequency-catheter ablation. Atrial function (minimal volume (LAmin), maximal volume (LAmax), stroke volume (SV), ejection fraction (EF)) was evaluated by two readers using a manual and an automatic tool and measurement time was evaluated. Results: Automated LA volume segmentation failed in one patient due to low LA enhancement (103HU). Mean LAmax, LAmin, SV and EF were 127.7 ml, 93 ml, 34.7 ml, 27.1% by the automated, and 122.7 ml, 89.9 ml, 32.8 ml, 26.3% by the manual method with no significant difference (p > 0.05) and high Pearsons correlation coefficients (r = 0.94, r = 0.94, r = 0.82 and r = 0.85, p < 0.0001), respectively. The automated method was significantly faster (p < 0.001). Interobserver variability was low for both methods with Pearson's correlation coefficients between 0.98 and 0.99 (p < 0.0001). Conclusions: Evaluation of LA volume and function with 64-slice MDCT is feasible with a very low interobserver variability. The automatic method is as accurate as the manual method but significantly less time consuming permitting a routine use in clinical practice before RF-catheter ablation.

  11. Student manual, Book 2: Orientation to occupational safety compliance in DOE

    Energy Technology Data Exchange (ETDEWEB)

    Colley, D.L.

    1993-10-01

    This is a student hand-book an Occupational Safety Compliance in DOE. Topics include the following: Electrical; materials handling & storage; inspection responsibilities & procedures; general environmental controls; confined space entry; lockout/tagout; office safety, ergonomics & human factors; medical & first aid, access to records; construction safety; injury/illness reporting system; and accident investigation procedures.

  12. Student manual, Book 2: Orientation to occupational safety compliance in DOE

    International Nuclear Information System (INIS)

    Colley, D.L.

    1993-01-01

    This is a student hand-book an Occupational Safety Compliance in DOE. Topics include the following: Electrical; materials handling ampersand storage; inspection responsibilities ampersand procedures; general environmental controls; confined space entry; lockout/tagout; office safety, ergonomics ampersand human factors; medical ampersand first aid, access to records; construction safety; injury/illness reporting system; and accident investigation procedures

  13. Orientation to occupational safety compliance in DOE: Student manual, Book 1

    International Nuclear Information System (INIS)

    1993-01-01

    The purpose of this document is to present Federal safety and health program requirements, provide an overview of DOE Safety and Health Programs, and to present information on the changes presently underway in the Department of Energy regarding safety and health. Public Law 91--596, Executive Order 12196,29 CFR 1960 and various related DOE Orders will be presented

  14. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1

    International Nuclear Information System (INIS)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety

  15. Nuclear Safety: Volume 29, No. 3: Technical progress review

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1988-07-01

    Nuclear Safety is a review journal that covers significant development in the field of nuclear safety. Its scope included the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Individual papers have been cataloged separately.

  16. Preliminary safety information document for the standard MHTGR. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-01-01

    This report contains information concerning: operational radionuclide control; occupational radiation protection, conduct of operations; initial test program; safety analysis; technical specifications; and quality assurance. (JDB)

  17. Energy Research and Development Administration, Division of Safety, Standards, and Compliance respirator manual

    International Nuclear Information System (INIS)

    Douglas, D.D.; Hack, A.L.; Held, B.J.; Revoir, W.H.

    1976-05-01

    The manual has been prepared to provide technical information for contractors of the Energy Research and Development Administration (ERDA) on the application of respiratory protective devices for protection against airborne contaminants, both radioactive and nonradioactive. The various elements of a respirator program including selection and maintenance of equipment and training of personnel are described to assist in establishing adequate programs

  18. Energy Research and Development Administration, Division of Safety, Standards, and Compliance respirator manual

    Energy Technology Data Exchange (ETDEWEB)

    Douglas, D.D.; Hack, A.L.; Held, B.J.; Revoir, W.H.

    1976-05-01

    The manual has been prepared to provide technical information for contractors of the Energy Research and Development Administration (ERDA) on the application of respiratory protective devices for protection against airborne contaminants, both radioactive and nonradioactive. The various elements of a respirator program including selection and maintenance of equipment and training of personnel are described to assist in establishing adequate programs.

  19. Forecasted electric power demands for the Baltimore Gas and Electric Company. Volume 1 and Volume 2. Documentation manual

    International Nuclear Information System (INIS)

    Estomin, S.L.; Beach, J.E.; Goldsmith, J.V.

    1991-05-01

    The two-volume report presents the results of an econometric forecast of peak load and electric power demand for the Baltimore Gas and Electric Company (BG ampersand E) through the year 2009. Separate energy sales models were estimated for residential sales in Baltimore City, residential sales in the BG ampersand E service area excluding Baltimore City, commercial sales, industrial sales, streetlighting sales, and Company use plus losses. Econometric equations were also estimated for electric space heating and air conditioning saturation in Baltimore City and in the remainder of the BG ampersand E service territory. In addition to the energy sales models and the electric space conditioning saturation models, econometric models of summer and winter peak demand on the BG ampersand E system were estimated

  20. Forecasted electric power demands for the Delmarva Power and Light Company. Volume 1 and Volume 2. Documentation manual

    International Nuclear Information System (INIS)

    Estomin, S.L.; Beach, J.E.

    1990-10-01

    The two-volume report presents the results of an econometric forecast of peak load and electric power demands for the Delmarva Power and Light Company (DP ampersand L) through the year 2008. Separate sets of models were estimated for the three jurisdictions served by DP ampersand L: Delaware, Maryland and Virginia. For both Delaware and Maryland, econometric equations were estimated for residential, commercial, industrial, and streetlighting sales. For Virginia, equations were estimated for residential, commercial plus industrial, and streetlighting sales; separate industrial and commercial equations were not estimated for Virginia due to the relatively small size of DP ampersand L's Virginia Industrial load. Wholesale sales were econometrically estimated for the DP ampersand L system as a whole. In addition to the energy sales models, an econometric model of annual (summer) peak demand was estimated for the Company

  1. STICAP: A linear circuit analysis program with stiff systems capability. Volume 1: Theory manual. [network analysis

    Science.gov (United States)

    Cooke, C. H.

    1975-01-01

    STICAP (Stiff Circuit Analysis Program) is a FORTRAN 4 computer program written for the CDC-6400-6600 computer series and SCOPE 3.0 operating system. It provides the circuit analyst a tool for automatically computing the transient responses and frequency responses of large linear time invariant networks, both stiff and nonstiff (algorithms and numerical integration techniques are described). The circuit description and user's program input language is engineer-oriented, making simple the task of using the program. Engineering theories underlying STICAP are examined. A user's manual is included which explains user interaction with the program and gives results of typical circuit design applications. Also, the program structure from a systems programmer's viewpoint is depicted and flow charts and other software documentation are given.

  2. Electric utility engineer`s FGD manual -- Volume 1: FGD process design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-04

    Part 1 of the Electric Utility Engineer`s Flue Gas Desulfurization (FGD) Manual emphasizes the chemical and physical processes that form the basis for design and operation of lime- and limestone-based FGD systems applied to coal- or oil-fired steam electric generating stations. The objectives of Part 1 are: to provide a description of the chemical and physical design basis for lime- and limestone-based wet FGD systems; to identify and discuss the various process design parameters and process options that must be considered in developing a specification for a new FGD system; and to provide utility engineers with process knowledge useful for operating and optimizing a lime- or limestone-based wet FGD system.

  3. ABAQUS-EPGEN: a general-purpose finite-element code. Volume 1. User's manual

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.

    1982-10-01

    This document is the User's Manual for ABAQUS/EPGEN, a general purpose finite element computer program, designed specifically to serve advanced structural analysis needs. The program contains very general libraries of elements, materials and analysis procedures, and is highly modular, so that complex combinations of features can be put together to model physical problems. The program is aimed at production analysis needs, and for this purpose aspects such as ease-of-use, reliability, flexibility and efficiency have received maximum attention. The input language is designed to make it straightforward to describe complicated models; the analysis procedures are highly automated with the program choosing time or load increments based on user supplied tolerances and controls; and the program offers a wide range of post-processing options for display of the analysis results

  4. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 2: user's manual (Revision 2)

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1985-07-01

    Revisions to the VIPRE code documents for Volume 2 are presented. These revisions conform to the changes made to VIPRE-01, CYCLE-00 to produce the new version of the code denoted by VIPRE-01, CYCLE-01. The first pages of the revisions specify where the replacement pages are to be inserted and which pages of the original documents should be retained

  5. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  6. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  7. Biosafety Manual

    Energy Technology Data Exchange (ETDEWEB)

    King, Bruce W.

    2010-05-18

    Work with or potential exposure to biological materials in the course of performing research or other work activities at Lawrence Berkeley National Laboratory (LBNL) must be conducted in a safe, ethical, environmentally sound, and compliant manner. Work must be conducted in accordance with established biosafety standards, the principles and functions of Integrated Safety Management (ISM), this Biosafety Manual, Chapter 26 (Biosafety) of the Health and Safety Manual (PUB-3000), and applicable standards and LBNL policies. The purpose of the Biosafety Program is to protect workers, the public, agriculture, and the environment from exposure to biological agents or materials that may cause disease or other detrimental effects in humans, animals, or plants. This manual provides workers; line management; Environment, Health, and Safety (EH&S) Division staff; Institutional Biosafety Committee (IBC) members; and others with a comprehensive overview of biosafety principles, requirements from biosafety standards, and measures needed to control biological risks in work activities and facilities at LBNL.

  8. Manual on Safety Aspects of the Design and Equipment of Hot Laboratories

    International Nuclear Information System (INIS)

    1969-01-01

    With the development of atomic energy application and research, hot laboratories are now being constructed in a number of countries. The present publication describes and discusses experience in several countries in designing equipment for these laboratories. The safe handling of highly radioactive substances is the main purpose of hot laboratory design and equipment. The manual aims at helping those persons, particularly in the developing countries, who plan to design and construct a new hot laboratory or modify an existing one. It does not deal in great detail with the engineering design of protective and handling equipment; these matters can be found in the comprehensive list of references. The manual itself covers only basic ideas and different approaches in the design of laboratory building, hot cells, shielded and glove boxes, fume cupboards, and handling and viewing equipment. Systems for transferring materials and main services are also discussed.

  9. Assembly, operation and disassembly manual for the Battelle Large Volume Water Sampler (BLVWS)

    International Nuclear Information System (INIS)

    Thomas, V.W.; Campbell, R.M.

    1984-12-01

    Assembly, operation and disassembly of the Battelle Large Volume Water Sampler (BLVWS) are described in detail. Step by step instructions of assembly, general operation and disassembly are provided to allow an operator completely unfamiliar with the sampler to successfully apply the BLVWS to his research sampling needs. The sampler permits concentration of both particulate and dissolved radionuclides from large volumes of ocean and fresh water. The water sample passes through a filtration section for particle removal then through sorption or ion exchange beds where species of interest are removed. The sampler components which contact the water being sampled are constructed of polyvinylchloride (PVC). The sampler has been successfully applied to many sampling needs over the past fifteen years. 9 references, 8 figures

  10. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes

  11. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  12. Hydride heat pump. Volume I. Users manual for HYCSOS system design program. [HYCSOS code

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, R.; Moritz, P.

    1978-05-01

    A method for the design and costing of a metal hydride heat pump for residential use and a computer program, HYCSOS, which automates that method are described. The system analyzed is one in which a metal hydride heat pump can provide space heating and space cooling powered by energy from solar collectors and electric power generated from solar energy. The principles and basic design of the system are presented, and the computer program is described giving detailed design and performance equations used in the program. The operation of the program is explained, and a sample run is presented. This computer program is part of an effort to design, cost, and evaluate a hydride heat pump for residential use. The computer program is written in standard Fortran IV and was run on a CDC Cyber 74 and Cyber 174 computer. A listing of the program is included as an appendix. This report is Volume 1 of a two-volume document.

  13. Rail Safety/Equipment Crashworthiness : Volume 3. Proposed Engineering Standards.

    Science.gov (United States)

    1978-07-01

    The document, the third of four volumes, contains recommended Engineering Standards prepared in the format of the standards published in the Code of Federal Regulations (Title 49, Transportation, Parts 200). The standards proposed provide improved oc...

  14. Nuclear safety, Volume 38, Number 1, January--March 1997

    Energy Technology Data Exchange (ETDEWEB)

    None

    1997-03-01

    This journal contains nine articles which fall under the following categories: (1) general safety considerations; (2) control and instrumentation; (3) design features (4) environmental effects; (5) US Nuclear Regulatory Commission information and analyses; and (6) recent developments.

  15. Unresolved safety issues summary. Aqua Book. Volume 6, No. 3

    International Nuclear Information System (INIS)

    Butts, J.

    1984-01-01

    The unresolved safety issues summary is designed to provide the management of the Nuclear Regulatory Commission with a quarterly overview of the progress and plans for completion of generic tasks addressing unresolved safety issues reported to Congress pursuant to Section 210 of the Energy Reorganization Act of 1974 as amended. This summary utilizes data collected from the Office of Nuclear Reactor Regulation, Office of Nuclear Regulatory Research, and the national laboratories and is prepared by the Office of Nuclear Reactor Regulation

  16. Leading Edge. Volume 7, Number 3. Systems Safety Engineering

    Science.gov (United States)

    2010-01-01

    foods were not always safe to eat given the sanitary conditions of the day. In 1943, the psychologist Abraham Maslow proposed a five-level... hierarchy of basic human needs, and safety was number two on this list. System safety is a specialized and formalized extension of our in- herent drive for...factors, hazards, mishaps, and ef- fects. The following is an example of each element within the hierarchy : An exposed sharp edge in a relay cabi- net

  17. Occupational Safety and Health Program Guidelines for Colleges and Universities. An Administrative Resource Manual.

    Science.gov (United States)

    Godbey, Frank W.; Hatch, Loren L.

    Designed as an aid for establishing and strengthening occupational safety and health programs on college and university campuses, this administrator guide is divided into four chapters. The first chapter defines and gives background information on the Occupational Safety and Health Act (OSHA). In addition, it presents a discussion of what the OSHA…

  18. Safety in Academic Chemistry Laboratories: Volume 2. Accident Prevention for Faculty and Administrators, 7th Edition.

    Science.gov (United States)

    American Chemical Society, Washington, DC.

    This book contains volume 2 of 2 and describes safety guidelines for academic chemistry laboratories to prevent accidents for college and university students. Contents include: (1) "Organizing for Accident Prevention"; (2) "Personal Protective Equipment"; (3) "Labeling"; (4) "Material Safety Data Sheets (MSDSs)"; (5) "Preparing for Medical…

  19. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences

  20. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  1. RELAP5/MOD3 code manual: User's guide and input requirements. Volume 2

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation

  2. Unresolved safety issues summary. Volume 3, Number 3. Aqua book

    International Nuclear Information System (INIS)

    1981-01-01

    The 'Unresolved Safety Issues' summary is designed to provide the management of the Nuclear Regulatory Commission with a quarterly overview of the progress and plans for completion of generic tasks addressing Unresolved Safety Issues reported to Congress pursuant to section 210 of The Energy Reorganization Act of 1974 as amended. This summary utilizes data collected from the Office of Nuclear Reactor Regulation, Office of Nuclear Regulatory Research, and the National Laboratories and is prepared by the office of Management and Program Analysis. The definition of what constitutes completion of an unresolved safety issue (USI) has recently been expanded to include the implementation of the technical resolution. This is in acknowledgement of the fact that real safety benefits occur only after the implementation has taken place. The schedules in this book include a milestone at the end of each action plan which represents the initiation of the implementation process both with respect to incorporation of the technical resolution in the NRC official guidance or requirements and also the application of changes to individual operating plants. The schedule for implementation will not normally be included in the task action plan(s) for the resolution of a USI since the nature and extent of the activities necessary to accomplish the implementation cannot normally be reasonably determined prior to the determination of a technical resolution. The progress and status for implementation of unresolved safety issues for which a technical resolution has been completed are reported specifically in a separate table provided in this summary

  3. Hippocampal volume change measurement: Quantitative assessment of the reproducibility of expert manual outlining and the automated methods FreeSurfer and FIRST

    NARCIS (Netherlands)

    Mulder, E.R.; de Jong, R.A.; Knol, D.L.; van Schijndel, R.A.; Cover, K.S.; Visser, P.J.; Barkhof, F.; Vrenken, H.

    2014-01-01

    Background: To measure hippocampal volume change in Alzheimer's disease (AD) or mild cognitive impairment (MCI), expert manual delineation is often used because of its supposed accuracy. It has been suggested that expert outlining yields poorer reproducibility as compared to automated methods, but

  4. National Household Education Surveys of 2003. Data File User's Manual, Volume II: Parent and Family Involvement in Education Survey. NCES 2004-102

    Science.gov (United States)

    Hagedorn, Mary; Montaquila, Jill; Vaden-Kiernan, Nancy; Kim, Kwang; Roth, Shelley Brock; Chapman, Christopher

    2004-01-01

    This manual provides documentation and guidance for users of the public-use data file for PFI-NHES: 2003. This volume contains a description of the content and organization of the data file, including useful information regarding questionnaire items and the various derived variables found on the file. Appended are the public-use data file layout,…

  5. Qualifications manual for Inspection and Enforcement personnel. Volume I. Detailed data sheets

    International Nuclear Information System (INIS)

    Bartley, H.J.; Heyer, F.H.K.; Schwartz, E.G.; Harrison, O.J.; Krass, I.W.; Welsh, D.M.

    1978-06-01

    This document is the General Research Corporation (GRC) final report on Task I Job/Position Qualifications Statements for the Headquarters and regional position of NRC/IE. It reflects the results of upgrading, expanding, and reformating the preliminary Task I report. To accomplish this, GRC made a thorough examination of all data developed during Task I, Task II (Career Fields), Task III (Training Evaluation), and Task IV (Hiring Qualifications), to include NRC/IE comments and management decisions received. The reformated Job/Position Qualifications Statements resulting from this effort (Appendix A) identify three job qualifications categories for each position--performance, attributes, and knowledge. Two formats are provided: a summarized one for manpower management in Volume I and a detailed one for career and training management in Volume II. In addition to recommending approval of the job/position qualifications, GRC proposes their use in conjunction with the hiring qualifications in the Task IV report to update NRC/IE job descriptions

  6. Planning manual for energy resource development on Indian lands. Volume IV. The environment

    Energy Technology Data Exchange (ETDEWEB)

    1978-03-01

    Many Indian tribes own rich deposits of very valuable energy resources. Existing and proposed uses of these tribal resources range from limited development of small oil and gas fields to large-scale extraction and conversion of coal, uranium, and oil shale. The adverse environmental impacts of such projects may create a conflict between a tribe's environmental policies and its economic, employment, and other long-term goals. The purpose of this volume is to provide tribal decision makers with reference documents on the mechanisms that are available to resolve such conflicts. This report focuses on the role of existing environmental laws in enabling tribes to achieve the needed balance among its objectives. Over a dozen major Federal statutes have been enacted to achieve this purpose. One law, the National Environmental Policy Act (NEPA), provides procedures to ensure that environmental factors are included in the Federal decision-making process. Numerous other laws, such as the Clean Air Act, have been enacted to prevent or control any negative environmental impacts of actual projects. This volume documents the key provisions of the laws and regulations, and discusses their effectiveness in meeting total needs. Also, tribal options to strengthen these mechanisms are highlighted. Sections II and III report on the role of NEPA in tribal development decisions. Section IV reviews those laws and regulations that control project operations.

  7. Nuclear safety. Volume 36, Number 2, July--December 1995

    Energy Technology Data Exchange (ETDEWEB)

    None

    1995-12-01

    The primary scope of the journal is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. The following subjects are covered here: (1) the Chernobyl accident; (2) general safety considerations; (3) accident analysis; (4) design features; (5) environmental effects; (6) operating experiences; (7) US NRC information and analyses; and (8) recent developments. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  8. The Effectiveness and Safety of Manual Therapy on Pain and Disability in Older Persons With Chronic Low Back Pain: A Systematic Review.

    Science.gov (United States)

    de Luca, Katie E; Fang, Sheng Hung; Ong, Justin; Shin, Ki-Soo; Woods, Samuel; Tuchin, Peter J

    2017-09-01

    The aim of this study was to perform a systematic review of the literature of the effectiveness and safety of manual therapy interventions on pain and disability in older persons with chronic low back pain (LBP). A literature search of 4 electronic databases was performed (PubMed, EMBASE, OVID, and CINAHL). Inclusion criteria included randomized controlled trials of manual therapy interventions on older persons who had chronic LBP. Effectiveness was determined by extracting and examining outcomes for pain and disability, with safety determined by the report of adverse events. The PEDro scale was used for quality assessment of eligible studies. The search identified 405 articles, and 38 full-text articles were assessed. Four studies met the inclusion criteria. All trials were of good methodologic quality and had a low risk of bias. The included studies provided moderate evidence supporting the use of manual therapy to reduce pain levels and alleviate disability. A limited number of studies have investigated the effectiveness and safety of manual therapy in the management of older people with chronic LBP. The current evidence to make firm clinical recommendations is limited. Research with appropriately designed trials to investigate the effectiveness and safety of manual therapy interventions in older persons with chronic LBP is required. Copyright © 2017. Published by Elsevier Inc.

  9. RELAP/MOD3 code manual: User's guidelines. Volume 5, Revision 1

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Schultz, R.R.

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume V contains guidelines that have solved over the past several years through the use of the RELAP5 code

  10. RELAP/MOD3 code manual: User`s guidelines. Volume 5, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; Schultz, R.R. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume V contains guidelines that have solved over the past several years through the use of the RELAP5 code.

  11. Rotorcraft Flight Simulation Computer Program C81 with Datamap Interface, Volume 2. Programmer’s Manual

    Science.gov (United States)

    1981-10-01

    overview of the computer program capabilities and the principal mathematical models incorporated in the program are given in Volume I of the documentation...blank card must still be placed in the appropriate place in the Model Data Set. For example, the mathematical model of a UH-lH would not need a wing...SIN %4114 LV OADs CC £6 $ 2. 27 4L 4+ 4 l’l <m L & % CC O 04i18K5 twsSTVEL 13L4.S2 ASK 2? b 41.815 . 8 ’ILbLS a CL 7 8 £0 1 £0 £0 12 £6 1 M90 A/ MUfI H

  12. Characterization of hazardous waste sites: a methods manual. Volume 2. Available sampling methods (second edition)

    International Nuclear Information System (INIS)

    Ford, P.J.; Turina, P.J.; Seely, D.E.

    1984-12-01

    Investigations at hazardous waste sites and sites of chemical spills often require on-site measurements and sampling activities to assess the type and extent of contamination. This document is a compilation of sampling methods and materials suitable to address most needs that arise during routine waste site and hazardous spill investigations. The sampling methods presented in this document are compiled by media, and were selected on the basis of practicality, economics, representativeness, compatability with analytical considerations, and safety, as well as other criteria. In addition to sampling procedures, sample handling and shipping, chain-of-custody procedures, instrument certification, equipment fabrication, and equipment decontamination procedures are described. Sampling methods for soil, sludges, sediments, and bulk materials cover the solids medium. Ten methods are detailed for surface waters, groundwater and containerized liquids; twelve are presented for ambient air, soil gases and vapors, and headspace gases. A brief discussion of ionizing radiation survey instruments is also provided

  13. Hippocampal Volume Is Reduced in Schizophrenia and Schizoaffective Disorder But Not in Psychotic Bipolar I Disorder Demonstrated by Both Manual Tracing and Automated Parcellation (FreeSurfer)

    Science.gov (United States)

    Arnold, Sara J. M.; Ivleva, Elena I.; Gopal, Tejas A.; Reddy, Anil P.; Jeon-Slaughter, Haekyung; Sacco, Carolyn B.; Francis, Alan N.; Tandon, Neeraj; Bidesi, Anup S.; Witte, Bradley; Poudyal, Gaurav; Pearlson, Godfrey D.; Sweeney, John A.; Clementz, Brett A.; Keshavan, Matcheri S.; Tamminga, Carol A.

    2015-01-01

    This study examined hippocampal volume as a putative biomarker for psychotic illness in the Bipolar-Schizophrenia Network on Intermediate Phenotypes (B-SNIP) psychosis sample, contrasting manual tracing and semiautomated (FreeSurfer) region-of-interest outcomes. The study sample (n = 596) included probands with schizophrenia (SZ, n = 71), schizoaffective disorder (SAD, n = 70), and psychotic bipolar I disorder (BDP, n = 86); their first-degree relatives (SZ-Rel, n = 74; SAD-Rel, n = 62; BDP-Rel, n = 88); and healthy controls (HC, n = 145). Hippocampal volumes were derived from 3Tesla T1-weighted MPRAGE images using manual tracing/3DSlicer3.6.3 and semiautomated parcellation/FreeSurfer5.1,64bit. Volumetric outcomes from both methodologies were contrasted in HC and probands and relatives across the 3 diagnoses, using mixed-effect regression models (SAS9.3 Proc MIXED); Pearson correlations between manual tracing and FreeSurfer outcomes were computed. SZ (P = .0007–.02) and SAD (P = .003–.14) had lower hippocampal volumes compared with HC, whereas BDP showed normal volumes bilaterally (P = .18–.55). All relative groups had hippocampal volumes not different from controls (P = .12–.97) and higher than those observed in probands (P = .003–.09), except for FreeSurfer measures in bipolar probands vs relatives (P = .64–.99). Outcomes from manual tracing and FreeSurfer showed direct, moderate to strong, correlations (r = .51–.73, P schizoaffective disorder, but not for psychotic bipolar I disorder, and may reflect a cumulative effect of divergent primary disease processes and/or lifetime medication use. Manual tracing and semiautomated parcellation regional volumetric approaches may provide useful outcomes for defining measurable biomarkers underlying severe mental illness. PMID:24557771

  14. 78 FR 48696 - Draft Revisions to the Marine Safety Manual, Volume III, Chapters 20-26

    Science.gov (United States)

    2013-08-09

    ... Commercial Vessel Compliance, U.S. Coast Guard; telephone 202-372-1208, email [email protected] . If...) clarifying roles, responsibilities, and facilitation of communications with the appropriate offices at Coast.... Jonathan C. Burton, Captain, U.S. Coast Guard, Director, Inspections and Compliance. [FR Doc. 2013-19359...

  15. Application of condition-based HRA method for a manual actuation of the safety features in a nuclear power Plant

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung-Cheol

    2006-01-01

    A practical approach to develop a more realistic fault-tree model with a consideration of various conditions endured by a human operator is proposed. In safety-critical systems, the generation failure of an actuation signal is caused by the concurrent failures of the automated systems and an operator action. These two sources of safety signals are complicatedly correlated. The failures of sensors or automated systems will cause a lack of necessary information for a human operator and result in error-forcing contexts such as the loss of corresponding alarms and indications. It is well known that the error-forcing contexts largely affect the operator's performance. An automated system which consists of multiple processing channels and complex components is also affected by the availability of the sensors. This paper proposes a condition-based human reliability assessment (CBHRA) method in order to address these complicated conditions in a practical way. We apply the CBHRA method to the manual actuation of the safety features such as a reactor trip and auxiliary feedwater actuation in Korean Standard Nuclear Power Plants. Even the human error probability of each given condition is simply assumed, the application results prove that the CBHRA effectively accommodates the complicated error-forcing contexts into the fault trees

  16. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  17. SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

    International Nuclear Information System (INIS)

    Davis, K.L.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made

  18. SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K.; Johnsen, E.C. [eds.; Allison, C.M. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported.

  19. SCDAP/RELAP5/MOD 3.1 Code Manual: Developmental assessment. Volume 5

    International Nuclear Information System (INIS)

    Hohorst, J.K.; Johnsen, E.C.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of Light Water Reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed code-to-data calculations performed using SCDAP/RELAP5/MOD3.1, as well as comparison calculations performed with earlier code versions. Results of full plant calculations which include Surry, TMI-2, and Browns Ferry are described. Results of a nodalization study, which accounted for both axial and radial nodalization of the core, are also reported

  20. SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Davis, K.L. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made.

  1. Planning manual for energy resource development on Indian lands. Volume III. Manpower and training

    Energy Technology Data Exchange (ETDEWEB)

    1978-03-01

    This volume addresses ways to bridge the gap between existing tribal skill levels and the skill levels required for higher-paying jobs in energy resource development projects. It addresses opportunities for technical, skilled, and semiskilled employment as well as professional positions, because it is important to have tribal participation at all levels of an operation. Section II, ''Energy-Related Employment Opportunities,'' covers three areas: (1) identification of energy-resource occupations; (2) description of these occupations; and (3) identification of skill requirements by type of occupation. Section III, ''Description of Training Programs,'' also covers three areas: (a) concept of a training-program model; (b) description of various training methods; and (c) an assessment of the cost of training, utilizing different programs. Section IV concentrates on development of a training program for target occupations, skills, and populations. Again this section covers three areas: (i) overview of the development of a skills training program; (ii) identification of target occupations, skills, and populations; and (iii) energy careers for younger tribal members.

  2. SCDAP/RELAP5/MOD 3.1 code manual: Interface theory. Volume 1

    International Nuclear Information System (INIS)

    Coryell, E.W.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of off-site power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume describes the organization and manner of the interface between severe accident models which are resident in the SCDAP portion of the code and hydrodynamic models which are resident in the RELAP5 portion of the code. A description of the organization and structure of SCDAP/RELAP5 is presented. Additional information is provided regarding the manner in which models in one portion of the code impact other parts of the code, and models which are dependent on and derive information from other subcodes

  3. Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

    International Nuclear Information System (INIS)

    Cottrell, W.B.; Passiakos, M.

    1982-06-01

    This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index

  4. Index to Nuclear Safety: a technical progress review by chrology, permuted title, and author, Volume 11(1) through Volume 20(6)

    Energy Technology Data Exchange (ETDEWEB)

    Cottrell, W B; Passiakos, M

    1980-06-01

    This index to Nuclear Safety, a bimonthly technical progress review, covers articles published in Nuclear Safety, Volume II, No. 1 (January-February 1970), through Volume 20, No. 6 (November-December 1979). It is divided into three sections: a chronological list of articles (including abstracts) followed by a permuted-title (KWIC) index and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center (NSIC), covers all safety aspects of nuclear power reactors and associated facilities. Over 600 technical articles published in Nuclear Safety in the last ten years are listed in this index.

  5. Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

    Energy Technology Data Exchange (ETDEWEB)

    Cottrell, W.B.; Passiakos, M.

    1982-06-01

    This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

  6. Hippocampal volume change measurement: quantitative assessment of the reproducibility of expert manual outlining and the automated methods FreeSurfer and FIRST.

    Science.gov (United States)

    Mulder, Emma R; de Jong, Remko A; Knol, Dirk L; van Schijndel, Ronald A; Cover, Keith S; Visser, Pieter J; Barkhof, Frederik; Vrenken, Hugo

    2014-05-15

    To measure hippocampal volume change in Alzheimer's disease (AD) or mild cognitive impairment (MCI), expert manual delineation is often used because of its supposed accuracy. It has been suggested that expert outlining yields poorer reproducibility as compared to automated methods, but this has not been investigated. To determine the reproducibilities of expert manual outlining and two common automated methods for measuring hippocampal atrophy rates in healthy aging, MCI and AD. From the Alzheimer's Disease Neuroimaging Initiative (ADNI), 80 subjects were selected: 20 patients with AD, 40 patients with mild cognitive impairment (MCI) and 20 healthy controls (HCs). Left and right hippocampal volume change between baseline and month-12 visit was assessed by using expert manual delineation, and by the automated software packages FreeSurfer (longitudinal processing stream) and FIRST. To assess reproducibility of the measured hippocampal volume change, both back-to-back (BTB) MPRAGE scans available for each visit were analyzed. Hippocampal volume change was expressed in μL, and as a percentage of baseline volume. Reproducibility of the 1-year hippocampal volume change was estimated from the BTB measurements by using linear mixed model to calculate the limits of agreement (LoA) of each method, reflecting its measurement uncertainty. Using the delta method, approximate p-values were calculated for the pairwise comparisons between methods. Statistical analyses were performed both with inclusion and exclusion of visibly incorrect segmentations. Visibly incorrect automated segmentation in either one or both scans of a longitudinal scan pair occurred in 7.5% of the hippocampi for FreeSurfer and in 6.9% of the hippocampi for FIRST. After excluding these failed cases, reproducibility analysis for 1-year percentage volume change yielded LoA of ±7.2% for FreeSurfer, ±9.7% for expert manual delineation, and ±10.0% for FIRST. Methods ranked the same for reproducibility of 1

  7. Fire Safety Aspects of Polymeric Materials. Volume 7. Buildings

    Science.gov (United States)

    1979-01-01

    Custodial Buildings 136 4.5.5 Retail Stores, Malls, etc. 138 l’ 4.5.6 Restaurants and Nightclubs 4.5.7 Public Assembly Occupancies - Auditoria , Theaters... auditoria , theaters, exhibition halls, arenas, transportation terminals; educational buildings and indus- trial buildings. Many of the fire safety...usage are developed. 4.5.7 Public Assembly Occupancies - Auditoria , Theaters, Exhibition Halls, Arenas, Transportation Terminals, Etc. The factors

  8. Radiological safety problems in intracavitary brachytherapy using the manual afterloading in Venezuela

    International Nuclear Information System (INIS)

    Lea, D.; Gonzales, E.; Gallardo, J.; Diaz, M.

    1996-01-01

    In three venezuelan public hospitals it was detected radioactive contamination for Cs-137 during inspection with the wipe test technique in areas where is applied intracavitary brachytherapy using manual afterloading. This caused a condition of great stress in the personal engaged the treatments. In each Cs-137 source in these hospitals was tested leakage. The source plastic holder was analysed. The authors were able to demonstrate: a) The radioactive contamination found in the treatment area had its origin in the Cs-137 capsule corrosion and b) The corrosion in the source capsule was produced by the radiolytic degradation of the PVC tube used as source-holder in the intracavitary treatments. (authors). 3 refs., 1 tab

  9. Health and Safety Research Division manual for the x-ray facility in Building 2008

    International Nuclear Information System (INIS)

    Stansbury, P.S.

    1977-11-01

    The facility in the east end of Building 2008, ORNL, consists of an x-ray machine and a shielded enclosure. The x-ray machine is of the constant potential type and can be operated continuously at generating potentials up to 125 kVcp with tube currents ranging from 0.1 μA to 10 mA. Both the generating potential and the tube current are highly regulated and stabilized. The machine is installed and operated within a large shielded enclosure. In addition to the x-ray machine and its ancillary equipment, the shielded enclosure contains many of the features of a general chemistry and physics laboratory including work benches, sinks, storage space, and electrical and gas service. This manual contains instructions for the safe operation of the x-ray machine

  10. Evaluation of left atrial function by multidetector computed tomography before left atrial radiofrequency-catheter ablation: Comparison of a manual and automated 3D volume segmentation method

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, Florian, E-mail: florian.wolf@meduniwien.ac.a [Department of Radiology, Medical University of Vienna, Vienna (Austria); Ourednicek, Petr [Philips Medical Systems, Prague (Czech Republic); Loewe, Christian [Department of Radiology, Medical University of Vienna, Vienna (Austria); Richter, Bernhard; Goessinger, Heinz David; Gwechenberger, Marianne [Department of Cardiology, Medical University of Vienna, Vienna (Austria); Plank, Christina; Schernthaner, Ruediger Egbert; Toepker, Michael; Lammer, Johannes [Department of Radiology, Medical University of Vienna, Vienna (Austria); Feuchtner, Gudrun M. [Department of Radiology, Innsbruck Medical University, Innsbruck (Austria); Institute of Diagnostic Radiology, University Hospital Zurich (Switzerland)

    2010-08-15

    Introduction: The purpose of this study was to compare a manual and automated 3D volume segmentation tool for evaluation of left atrial (LA) function by 64-slice multidetector-CT (MDCT). Methods and materials: In 33 patients with paroxysmal atrial fibrillation a MDCT scan was performed before radiofrequency-catheter ablation. Atrial function (minimal volume (LAmin), maximal volume (LAmax), stroke volume (SV), ejection fraction (EF)) was evaluated by two readers using a manual and an automatic tool and measurement time was evaluated. Results: Automated LA volume segmentation failed in one patient due to low LA enhancement (103HU). Mean LAmax, LAmin, SV and EF were 127.7 ml, 93 ml, 34.7 ml, 27.1% by the automated, and 122.7 ml, 89.9 ml, 32.8 ml, 26.3% by the manual method with no significant difference (p > 0.05) and high Pearsons correlation coefficients (r = 0.94, r = 0.94, r = 0.82 and r = 0.85, p < 0.0001), respectively. The automated method was significantly faster (p < 0.001). Interobserver variability was low for both methods with Pearson's correlation coefficients between 0.98 and 0.99 (p < 0.0001). Conclusions: Evaluation of LA volume and function with 64-slice MDCT is feasible with a very low interobserver variability. The automatic method is as accurate as the manual method but significantly less time consuming permitting a routine use in clinical practice before RF-catheter ablation.

  11. Risk monitor - a tool for operational safety assessment risk monitor - user's manual

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Vinod, Gopika; Saraf, R.K.; Ghosh, A.K.

    2006-06-01

    Probabilistic Safety Assessment has become a key tool as on today to identify and understand Nuclear Power Plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. Risk Monitor is a PC based tool, which computes the real time safety level and assists plant personnel to manage day-to-day activities. Risk Monitor is a PC based user friendly software tool used for modification and re-analysis of a nuclear Power plant. Operation of Risk Monitor is based on PSA methods for assisting in day to day applications. Risk Monitoring programs can assess the risk profile and are used to optimize the operation of Nuclear Power Plants with respect to a minimum risk level over the operating time. This report presents the background activities of Risk Monitor, its application areas and the step by step procedure for the user.to interact with the software. This software can be used with the PSA model of any Nuclear Power Plant. (author)

  12. 19th JANNAF Safety and Environmental Protection Subcommittee Meeting. Volume 1

    Science.gov (United States)

    Cocchiaro, J. E. (Editor); Becker, D. L. (Editor)

    2002-01-01

    This volume, the first of two volumes, is a compilation of 22 unclassified/unlimited technical papers presented at the 19th Joint Army-Navy-NASA-Air Force (JANNAF) Safety & Environmental Protection Subcommittee Meeting. The meeting was held 18-21 March 2002 at the Sheraton Colorado Springs Hotel, Colorado Springs, Colorado. Topics covered include green energetic materials and life cycle pollution prevention; space launch range safety; propellant/munitions demilitarization, recycling, and reuse: and environmental and occupational health aspects of propellants and energetic materials.

  13. Inventory of Federal energy-related environment and safety research for FY 1977. Volume II. Project listings

    Energy Technology Data Exchange (ETDEWEB)

    1978-07-01

    This volume contains Biomedical and Environmental Research, Environmental Control Technology Research, and Operational and Environmental Safety Research project listings. The projects are ordered numerically by log number.

  14. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  15. MELCOR computer code manuals

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  16. GASSAR-6 (General Atomic Standard Safety Analysis Report). Volume 1

    International Nuclear Information System (INIS)

    1975-01-01

    A standard nuclear steam system for a 3000 MW(t), 1160 MW(e) high temperature gas-cooled reactor (HTGR) nuclear power station is described. The HTGR operates on a uranium-235/thorium-232 cycle. Spherical fuel pellets are coated with multiple layers of pyrolytic carbon, bonded into rods, and encased in hexagonal graphite fuel elements. The core is nuclear-purity-grade, near isotropic graphite machined in hexagonal blocks, which serves as moderator and the heat transfer medium between the fuel and the coolant. Forced helium is the primary coolant and water is the secondary coolant. A prestressed concrete reactor vessel (PCRV) houses the reactor core, primary coolant system and portions of the secondary coolant system, steam generators and circulators. A continuous internal steel liner in the PCRV acts as the primary coolant boundary and sealing membrane. The control rods contain boron carbide in a graphite matrix sheathed in Incoloy 800 cans. Boron constitutes about 40 percent of the absorber material by volume. The control rod drives provide insertion and withdrawal rates consistent with the required reactivity changes for operational load fluctuations and reactor shutdown. Control rods have shim and trip capability. (U.S.)

  17. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  18. Waste Isolation Pilot Plant Safety Analysis Report. Volume 5

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  19. Waste Isolation Pilot Plant Safety Analysis Report. Volume 1

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection: Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating control and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  20. Waste Isolation Pilot Plant Safety Analysis Report. Volume 4

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  1. Waste Isolation Pilot Plant Safety Analysis Report. Volume 2

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Analysis Report (SAR) has been prepared by the US Department of Energy (DOE) to support the construction and operation of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP facility is designed to receive, inspect, emplace, and store unclassified defense-generated transuranic wastes in a retrievable fashion in an underground salt medium and to conduct studies and perform experiments in salt with high-level wastes. Upon the successful completion of these studies and experiments, WIPP is designed to serve as a permanent facility. The first chapter of this report provides a summary of the location and major design features of WIPP. Chapters 2 through 5 describe the site characteristics, design criteria, and design bases used in the design of the plant and the plant operations. Chapter 6 discusses radiation protection; Chapters 7 and 8 present an accident analysis of the plant and an assessment of the long-term waste isolation at WIPP. The conduct of operations and operating controls and limits are discussed in Chapters 9 and 10. The quality assurance programs are described in Chapter 11

  2. Sodium fast reactor safety and licensing research plan - Volume I

    International Nuclear Information System (INIS)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-01-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  3. NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation

    Science.gov (United States)

    Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Smith, Curtis; Stamatelatos, Michael; Youngblood, Robert

    2011-01-01

    System safety assessment is defined in NPR 8715.3C, NASA General Safety Program Requirements as a disciplined, systematic approach to the analysis of risks resulting from hazards that can affect humans, the environment, and mission assets. Achievement of the highest practicable degree of system safety is one of NASA's highest priorities. Traditionally, system safety assessment at NASA and elsewhere has focused on the application of a set of safety analysis tools to identify safety risks and formulate effective controls.1 Familiar tools used for this purpose include various forms of hazard analyses, failure modes and effects analyses, and probabilistic safety assessment (commonly also referred to as probabilistic risk assessment (PRA)). In the past, it has been assumed that to show that a system is safe, it is sufficient to provide assurance that the process for identifying the hazards has been as comprehensive as possible and that each identified hazard has one or more associated controls. The NASA Aerospace Safety Advisory Panel (ASAP) has made several statements in its annual reports supporting a more holistic approach. In 2006, it recommended that "... a comprehensive risk assessment, communication and acceptance process be implemented to ensure that overall launch risk is considered in an integrated and consistent manner." In 2009, it advocated for "... a process for using a risk-informed design approach to produce a design that is optimally and sufficiently safe." As a rationale for the latter advocacy, it stated that "... the ASAP applauds switching to a performance-based approach because it emphasizes early risk identification to guide designs, thus enabling creative design approaches that might be more efficient, safer, or both." For purposes of this preface, it is worth mentioning three areas where the handbook emphasizes a more holistic type of thinking. First, the handbook takes the position that it is important to not just focus on risk on an individual

  4. Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank

    International Nuclear Information System (INIS)

    Scofield, N.R.; Hardy, H.A.; Laats, E.T.

    1983-02-01

    The United States Nuclear Regulatory Commission (NRC) has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water safety research programs. Local direction of the program is provided by E G and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the NRC. The NRC/DAE Data Bank and the capabilities of the data access software are described

  5. Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank

    International Nuclear Information System (INIS)

    Scofield, N.R.; Hardy, H.A.; Laats, E.T.

    1984-03-01

    The United States NRC has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. Local direction of the program is provided by EG and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the NRC. The NRC/DAE Data Bank and the capabilities of the data access software are described in this document

  6. Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank. Revision 3

    International Nuclear Information System (INIS)

    Hardy, H.A.; Laats, E.T.

    1985-03-01

    The United States Nuclear Regulatory Commission (USNRC) has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. Local direction of the program is provided by EG and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data to be used by code development and assessment groups in meeting the code correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the USNRC. The NRC/DAE Data Bank and the capabilities of the data access software are described in this document

  7. Fast reactor safety: proceedings of the international topical meeting. Volume 2

    International Nuclear Information System (INIS)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments

  8. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    Energy Technology Data Exchange (ETDEWEB)

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  9. Safety work organization in nuclear power plant. A9. Volume 2

    International Nuclear Information System (INIS)

    1985-01-01

    The second volume provides the laws, directives, major standards, principles, lists of selected workplaces where woman work is prohibited, instructions for new personnel, general principles of workplace safety, reports and provisions by commissions for reporting accidents and injuries, recourses, etc. (J.P.)

  10. Sensitivity and uncertainty analyses applied to criticality safety validation. Volume 2

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

    1999-01-01

    This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies developed in Volume 1 to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the existing S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently in use by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The methods for application of S/U and generalized linear-least-square methodology (GLLSM) tools to the criticality safety validation procedures were described in Volume 1 of this report. Volume 2 of this report presents the application of these procedures to the validation of criticality safety analyses supporting uranium operations where enrichments are greater than 5 wt %. Specifically, the traditional k eff trending analyses are compared with newly developed k eff trending procedures, utilizing the D and c k coefficients described in Volume 1. These newly developed procedures are applied to a family of postulated systems involving U(11)O 2 fuel, with H/X values ranging from 0--1,000. These analyses produced a series of guidance and recommendations for the general usage of these various techniques. Recommendations for future work are also detailed

  11. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs.

  12. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User's manual

    International Nuclear Information System (INIS)

    1995-01-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs

  13. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Safety of nuclear fuel cycle facilities. Vol. 9 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2008-11-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. The INPRO manual is comprised of an overview volume (No. 1), and eight additional volumes covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of nuclear reactors (Volume 8), and safety of nuclear fuel cycle facilities (laid out in this report) (Volume 9).This report elaborates on the guidance given in the INPRO report 'Methodology for the assessment of innovative nuclear reactors and fuel cycles', IAEA-TECDOC-1434, and the previous INPRO report 'Guidance for the evaluation for innovative nuclear reactors and fuel cycles', IAEA-TECDOC-1362 (2003), in the area of safety of nuclear reactors. The present version of this manual deals with safety issues related to design and operation of mining, milling, refining, conversion, enrichment, fuel fabrication, fuel storage and fuel reprocessing facilities. The INPRO Manual starts with an introduction in Chapter 1. Chapter 2 sets out the necessary input for an INPRO assessment of the safety of an innovative nuclear fuel cycle facility. This includes information on the design for the plant and the safety

  14. Inventory of Federal energy-related environment and safety research for FY 1978. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    1979-12-01

    The FY 1978 Federal Inventory is a compilation of 3225 federally funded energy-related environmental and safety reserch projects. It consists of three volumes: an executive summary providing an overview of the data (Volume I), a catalog listing each Inventory project followed by series of indexes (Volume II), and an interactive terminal guide giving instructions for on-line data retrieval (Volume III). Volume I reviews the inventory data as a whole and also within each of three major categories: biomedical and environmental research, environmental control technology research, and operational safety research

  15. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  16. A fast reactor transient analysis methodology for PCs: Volume 3, LTC program manual of the QuickBASIC code

    International Nuclear Information System (INIS)

    Ott, K.O.; Chung, L.

    1992-06-01

    This manual augments the detailed manual of the GW-BASIC version of the LTC code for an application in QuickBASIC. As most of the GW-BASIC coding of this program for ''LMR Transient Calculations'' is compatible with QuickBASIC, this manual pertains primarily to the required changes, such as the handling of input and output. The considerable reduction in computation time achieved by this conversion is demonstrated for two sample problems, using a variety of hardware and execution options. The revised code is listed. Although the severe storage limitations of GW-BASIC no longer apply, the LOF transient path has not been completed in this QuickBASIC code. Its advantages are thus primarily in the much faster running time for TOP and LOHS transients. For the fastest PC hardware (486) and execution option the computation time is reduced by a factor of 124 compared to GW-BASIC on a 386/20

  17. 2009 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    2009-01-01

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety (safety organisation, general measures, measures related to various risks, inspections, control of emergency situations, audits and second level control, measures in basic nuclear installations) and radioprotection (organisation, significant facts, dosimeter results). It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of measurements of releases and of their impact on the environment (chemical and radiological assessment). Then after a description of measures to limit the volume of stored radioactive wastes and their impact on health and on the environment, tables indicate the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation) for which a significant event occurred on the 6 October 2009. For these installations, release measurements concern gaseous and liquid releases

  18. 2007 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    2007-01-01

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety: safety organisation, general measures, measures related to various risks, control of emergency situations, inspections, audits and second level control, measures in basic nuclear installations. It describes the measures concerning the radioprotection on this site: organisation, significant facts, and dose measurement results. It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of release measurements and of radiological and chemical assessments of the impact these releases on the environment. The report then describes measures implemented to limit the volume of stored radioactive wastes and also their impact on health and on the environment. It provides a series of tables indicating the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. It reports the recommendations expressed by the CHSCT (committee on hygiene, security and working conditions) after the 2006 report. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation), for which the significant events occurred on the 13. of March and on the 25. of May 2007. For these installations, release measurements concern gaseous and liquid releases

  19. New highway accident location manual for Missouri.

    Science.gov (United States)

    2013-12-01

    The Missouri HAL manual is used to identify, analyze, and correct high crash locations, and has not been updated since : 1999. This new edition brings the manual up to date, while incorporating the methodology of the national Highway Safety : Manual ...

  20. Site Environmental Report for 2006. Volume I, Environment, Health, and Safety Division

    Energy Technology Data Exchange (ETDEWEB)

    None

    2007-09-30

    Each year, Ernest Orlando Lawrence Berkeley National Laboratory prepares an integrated report on its environmental programs to satisfy the requirements of United States Department of Energy Order 231.1A, Environment, Safety, and Health Reporting.1 The Site Environmental Report for 2006 summarizes Berkeley Lab’s environmental management performance, presents environmental monitoring results, and describes significant programs for calendar year 2006. (Throughout this report, Ernest Orlando Lawrence Berkeley National Laboratory is referred to as “Berkeley Lab,” “the Laboratory,” “Lawrence Berkeley National Laboratory,” and “LBNL.”) The report is separated into two volumes. Volume I is organized into an executive summary followed by six chapters that contain an overview of the Laboratory, a discussion of the Laboratory’s environmental management system, the status of environmental programs, and summarized results from surveillance and monitoring activities. Volume II contains individual data results from surveillance and monitoring activities.

  1. The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

    International Nuclear Information System (INIS)

    2015-08-01

    Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

  2. NASTRAN thermal analyzer: Theory and application including a guide to modeling engineering problems, volume 1. [thermal analyzer manual

    Science.gov (United States)

    Lee, H. P.

    1977-01-01

    The NASTRAN Thermal Analyzer Manual describes the fundamental and theoretical treatment of the finite element method, with emphasis on the derivations of the constituent matrices of different elements and solution algorithms. Necessary information and data relating to the practical applications of engineering modeling are included.

  3. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  4. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  5. Efficacy and Safety of Manual Partial Red Cell Exchange in the Management of Severe Complications of Sickle Cell Disease in a Developing Country

    Directory of Open Access Journals (Sweden)

    B. F. Faye

    2017-01-01

    Full Text Available Introduction. The realization of red cell exchange (RCE in Africa faces the lack of blood, transfusion safety, and equipment. We evaluated its efficacy and safety in severe complications of sickle cell disease. Patients and Method. Manual partial RCE was performed among sickle cell patients who had severe complications. Efficacy was evaluated by clinical evolution, blood count, and electrophoresis of hemoglobin. Safety was evaluated on adverse effects, infections, and alloimmunization. Results. We performed 166 partial RCE among 44 patients including 41 homozygous (SS and 2 heterozygous composites SC and 1 S/β0-thalassemia. The mean age was 27.9 years. The sex ratio was 1.58. The regression of symptoms was complete in 100% of persistent vasoocclusive crisis and acute chest syndrome, 56.7% of intermittent priapism, and 30% of stroke. It was partial in 100% of leg ulcers and null in acute priapism. The mean variations of hemoglobin and hematocrit rate after one procedure were, respectively, +1.4 g/dL and +4.4%. That of hemoglobin S after 2 consecutive RCE was −60%. Neither alloimmunization nor viral seroconversion was observed. Conclusion. This work shows the feasibility of manual partial RCE in a low-resource setting and its efficacy and safety during complications of SCD outside of acute priapism.

  6. Efficacy and Safety of Manual Partial Red Cell Exchange in the Management of Severe Complications of Sickle Cell Disease in a Developing Country

    Science.gov (United States)

    Sow, D.; Seck, M.; Dieng, N.; Toure, S. A.; Gadji, M.; Senghor, A. B.; Gueye, Y. B.; Sy, D.; Sall, A.; Dieye, T. N.; Toure, A. O.; Diop, S.

    2017-01-01

    Introduction The realization of red cell exchange (RCE) in Africa faces the lack of blood, transfusion safety, and equipment. We evaluated its efficacy and safety in severe complications of sickle cell disease. Patients and Method Manual partial RCE was performed among sickle cell patients who had severe complications. Efficacy was evaluated by clinical evolution, blood count, and electrophoresis of hemoglobin. Safety was evaluated on adverse effects, infections, and alloimmunization. Results We performed 166 partial RCE among 44 patients including 41 homozygous (SS) and 2 heterozygous composites SC and 1 S/β0-thalassemia. The mean age was 27.9 years. The sex ratio was 1.58. The regression of symptoms was complete in 100% of persistent vasoocclusive crisis and acute chest syndrome, 56.7% of intermittent priapism, and 30% of stroke. It was partial in 100% of leg ulcers and null in acute priapism. The mean variations of hemoglobin and hematocrit rate after one procedure were, respectively, +1.4 g/dL and +4.4%. That of hemoglobin S after 2 consecutive RCE was −60%. Neither alloimmunization nor viral seroconversion was observed. Conclusion This work shows the feasibility of manual partial RCE in a low-resource setting and its efficacy and safety during complications of SCD outside of acute priapism. PMID:28584527

  7. Intra- and interobserver variability of MRI-based volume measurements of the hippocampus and amygdala using the manual ray-tracing method

    International Nuclear Information System (INIS)

    Achten, E.; Deblaere, K.; Damme, F. van; Kunnen, M.; Wagter, C. de; Boon, P.; Reuck, J. de

    1998-01-01

    We studied the intra- and interobserver variability of volume measurments of the hippocampus (HC) and the amygdala as applied to the detection of HC atrophy in patients with complex partial seizures (CPE), measuring the volumes of the HC and amygdala of 11 normal volunteers and 12 patients with presumed CPE, using the manual ray-tracing method. Two independent observers performed these measurements twice each using home-made software. The intra- and interobserver variability of the absolute volumes and of the normalised left-to-right volume differences (δV) between the HC (δV HC ), the amygdala (δV A ) and the sum of both (δV HCA) were assessed. In our mainly right-handed normals, the right HC and amygdala were on average 0.05 and 0.03 ml larger respectively than on the left. The interobserver variability for volume measurements in normal subjects was 1.80 ml for the HC and 0.82 ml for the amygdala, the intraobserver variability roughly one third of these values. The interobserver variability coefficient in normals was 3.6 % for δV HCA , 4.7 % for δV HC and 7.3 % for δV A . The intraobserver variability coefficient was 3.4 % for δV HCA , 4.2 % for δV HC amd 5.6 % for δV A . The variability in patients was the same for volume differences less than 5 % either side of the interval for normality, but was higher when large volume differences were encountered, is probably due to the lack of thresholding and/or normalisation. Cutoff values for lateralisation with the δV were defined. No intra- or interobserver lateralisation differences were encountered with δV HCA and δV HC . From these observations we conclude that the manual ray-tracing method is a robust method for lateralisation in patients with TLE. Due to its higher variability, this method is less suited to measure absolute volumes. (orig.) (orig.)

  8. Intra- and interobserver variability of MRI-based volume measurements of the hippocampus and amygdala using the manual ray-tracing method

    Energy Technology Data Exchange (ETDEWEB)

    Achten, E.; Deblaere, K.; Damme, F. van; Kunnen, M. [MR Department 1K12, University Hospital Gent (Belgium); Wagter, C. de [Department of Radiotherapy and Nuclear Medicine, University Hospital Gent (Belgium); Boon, P.; Reuck, J. de [Department of Neurology, University Hospital Gent (Belgium)

    1998-09-01

    We studied the intra- and interobserver variability of volume measurments of the hippocampus (HC) and the amygdala as applied to the detection of HC atrophy in patients with complex partial seizures (CPE), measuring the volumes of the HC and amygdala of 11 normal volunteers and 12 patients with presumed CPE, using the manual ray-tracing method. Two independent observers performed these measurements twice each using home-made software. The intra- and interobserver variability of the absolute volumes and of the normalised left-to-right volume differences ({delta}V) between the HC ({delta}V{sub HC}), the amygdala ({delta}V{sub A}) and the sum of both ({delta}V{sub HCA)} were assessed. In our mainly right-handed normals, the right HC and amygdala were on average 0.05 and 0.03 ml larger respectively than on the left. The interobserver variability for volume measurements in normal subjects was 1.80 ml for the HC and 0.82 ml for the amygdala, the intraobserver variability roughly one third of these values. The interobserver variability coefficient in normals was 3.6 % for {delta}V{sub HCA}, 4.7 % for {delta}V{sub HC} and 7.3 % for {delta}V{sub A}. The intraobserver variability coefficient was 3.4 % for {delta}V{sub HCA}, 4.2 % for {delta}V{sub HC} amd 5.6 % for {delta}V{sub A}. The variability in patients was the same for volume differences less than 5 % either side of the interval for normality, but was higher when large volume differences were encountered, is probably due to the lack of thresholding and/or normalisation. Cutoff values for lateralisation with the {delta}V were defined. No intra- or interobserver lateralisation differences were encountered with {delta}V{sub HCA} and {delta}V{sub HC}. From these observations we conclude that the manual ray-tracing method is a robust method for lateralisation in patients with TLE. Due to its higher variability, this method is less suited to measure absolute volumes. (orig.) (orig.) With 2 figs., 7 tabs., 23 refs.

  9. Safety and feasibility of atrial fibrillation ablation using Amigo® system versus manual approach: A pilot study

    Directory of Open Access Journals (Sweden)

    Antonio Scarà

    2018-03-01

    Full Text Available Background: The Amigo® Remote Catheter System is a relatively new robotic system for catheter navigation. This study compared feasibility and safety using Amigo (RCM versus manual catheter manipulation (MCM to treat paroxysmal atrial fibrillation (PAF. Contact force (CF and force-time integral (FTI values obtained during pulmonary vein isolation (PVI ablation were compared. Methods: Forty patients were randomly selected for either RCM (20 or MCM (20. All were studied with the Thermocool® SmartTouch® force-sensing catheter (STc. Contact Force (CF, Force Time Integral (FTI and procedure-related data, were measured/stored in the CARTO®3. Results: All cases achieved complete PVI without major complications. Mean CF was significantly higher in the RCM group (13.3 ± 7.7 g in RCM vs. 12.04 ± 7.42 g in MCM p < 0.001, as was overall mean FTI (425.6 gs ± 199.6 gs with RCM and 407.5 gs ± 288.0 gs in MCM (p = 0.007 and was more likely to fall into the optimal FTI range (400-1000 using RCM (66.1% versus 49.1%, p < 0.001. FTI was significantly more likely to fall within the optimal range in each PV, as was CF within its optimal range in the right PVs, but trended higher in the left PVs. Freedom from atrial tachyarrhythmia was 90.0% for the RCM and 70.0% for the MCM group (p = 0,12 at 540 days follow-up. Conclusions: This pilot study suggests that use of the Amigo RCM system, with STc catheter, seems to be safe and effective for PVI ablation in paroxysmal AF patients. A not statistically significant favorable trend was observed for RCM in term of AF-free survival. Keywords: Atrial fibrillation, Catheter ablation, Remote robotic ablation

  10. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K. J.

    1979-01-01

    Volume 2 contains appendixes on small MOX fuel fabrication facility description, site description, residual radionuclide inventory estimates, decommissioning, financing, radiation dose methodology, general considerations, packaging and shipping of radioactive materials, cost assessment, and safety (JRD)

  11. On the volume of cremated remains - a comparative study of archaeologically recovered cremated bone volume as measured manually and assessed by Computed Tomography and by Stereology

    DEFF Research Database (Denmark)

    Harvig, L.; Lynnerup, Niels

    2013-01-01

    Weight but occasionally also the volume of cremated human remains may often be the primary basis for interpreting the many stages involved in past cremation ceremonies. However, methods used for describing volume for cremated remains are extremely varying and biased by many factors. Here we evalu...

  12. MORT User's Manual for use with the Management Oversight and Risk Tree analytical logic diagram. [Contains a list of System Safety Development Center publications

    Energy Technology Data Exchange (ETDEWEB)

    Knox, N.W.; Eicher, R.W.

    1992-02-01

    This report contains the User's Manual for MORT (Management Oversight and Risk Tree), a logic diagram in the form of a work sheet'' that illustrates a long series of interrelated questions. MORT is a comprehensive analytical procedure that provides a disciplined method for determining the causes and contributing factors of major accidents. Alternatively, it serves as a tool to evaluate the quality of an existing system. While similar in many respects to fault tree analysis, MORT is more generalized and presents over 1,500 specific elements of an ideal universal'' management program for optimizing environment, safety and health, and other programs. This User's Manual is intended to be used with the MORT diagram dated February 1992.

  13. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist.

  14. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1998-05-01

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist

  15. Impacts of Outer Continental Shelf (OCS) development on recreation and tourism. Volume 4. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    1987-04-01

    The final report for the project is presented in five volumes. The project sought to determine the impact of Outer Continental Shelf development on recreation and tourism in California. This volume is the User's Guide. It includes the following topics: Introduction and Summary Guide; Input Data Files; Gravity Model Programs; Economic Effects Model Programs; Consumer Surplus Model Programs; References; and Appendices.

  16. Safety in Academic Chemistry Laboratories: Volume 1. Accident Prevention for College and University Students, 7th Edition.

    Science.gov (United States)

    American Chemical Society, Washington, DC.

    This book contains volume 1 of 2 and describes safety guidelines for academic chemistry laboratories to prevent accidents for college and university students. Contents include: (1) "Your Responsibility for Accident Prevention"; (2) "Guide to Chemical Hazards"; (3) "Recommended Laboratory Techniques"; and (4) "Safety Equipment and Emergency…

  17. Kilowatt isotope power system. Phase II plan. Volume V. Safety, quality assurance and reliability

    International Nuclear Information System (INIS)

    1978-01-01

    The development of a Kilowatt Isotope Power System (KIPS) was begun in 1975 for the purpose of satisfying the power requirements of satellites in the 1980's. The KIPS is a 238 PuO 2 -fueled organic Rankine cycle turbine power system to provide a design output of 500 to 2000 W. Included in this volume are: launch and flight safety considerations; quality assurance techniques and procedures to be followed through system fabrication, assembly and inspection; and the reliability program made up of reliability prediction analysis, failure mode analysis and criticality analysis

  18. Probabilistic safety analysis procedures guide, Sections 8-12. Volume 2, Rev. 1

    International Nuclear Information System (INIS)

    McCann, M.; Reed, J.; Ruger, C.; Shiu, K.; Teichmann, T.; Unione, A.; Youngblood, R.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. The first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. This second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  19. Probabilistic safety analysis procedures guide. Sections 1-7 and appendices. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Bari, R.A.; Buslik, A.J.; Cho, N.Z.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. This first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. The second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  20. Guidelines for the verification and validation of expert system software and conventional software: User`s manual. Volume 7

    Energy Technology Data Exchange (ETDEWEB)

    Mirsky, S.M.; Hayes, J.E.; Miller, L.A. [Science Applications International Corp., McLean, VA (United States)

    1995-03-01

    This report provides a step-by-step guide, or user manual, for personnel responsible for the planning and execution of the verification and validation (V&V), and developmental testing, of expert systems, conventional software systems, and various other types of artificial intelligence systems. While the guide was developed primarily for applications in the utility industry, it applies well to all industries. The user manual has three sections. In Section 1 the user assesses the stringency of V&V needed for the system under consideration, identifies the development stage the system is in, and identifies the component(s) of the system to be tested next. These three pieces of information determine which Guideline Package of V&V methods is most appropriate for those conditions. The V&V Guideline Packages are provided in Section 2. Each package consists of an ordered set of V&V techniques to be applied to the system, guides on choosing the review/evaluation team, measurement criteria, and references to a book or report which describes the application of the method. Section 3 presents details of 11 of the most important (or least well-explained in the literature) methods to assist the user in applying these techniques accurately.

  1. Guidelines for the verification and validation of expert system software and conventional software: User's manual. Volume 7

    International Nuclear Information System (INIS)

    Mirsky, S.M.; Hayes, J.E.; Miller, L.A.

    1995-03-01

    This report provides a step-by-step guide, or user manual, for personnel responsible for the planning and execution of the verification and validation (V ampersand V), and developmental testing, of expert systems, conventional software systems, and various other types of artificial intelligence systems. While the guide was developed primarily for applications in the utility industry, it applies well to all industries. The user manual has three sections. In Section 1 the user assesses the stringency of V ampersand V needed for the system under consideration, identifies the development stage the system is in, and identifies the component(s) of the system to be tested next. These three pieces of information determine which Guideline Package of V ampersand V methods is most appropriate for those conditions. The V ampersand V Guideline Packages are provided in Section 2. Each package consists of an ordered set of V ampersand V techniques to be applied to the system, guides on choosing the review/evaluation team, measurement criteria, and references to a book or report which describes the application of the method. Section 3 presents details of 11 of the most important (or least well-explained in the literature) methods to assist the user in applying these techniques accurately

  2. RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

    International Nuclear Information System (INIS)

    Moore, R.L.; Sloan, S.M.; Schultz, R.R.; Wilson, G.E.

    1996-10-01

    Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available

  3. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 5.0: Data loading manual. Volume 10

    International Nuclear Information System (INIS)

    VanHorn, R.L.; Wolfram, L.M.; Fowler, R.D.; Beck, S.T.; Smith, C.L.

    1995-04-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) suite of programs can be used to organize and standardize in an electronic format information from probabilistic risk assessments or individual plant examinations. The Models and Results Database (MAR-D) program of the SAPHIRE suite serves as the repository for probabilistic risk assessment and individual plant examination data and information. This report demonstrates by examples the common electronic and manual methods used to load these types of data. It is not a stand alone document but references documents that contribute information relative to the data loading process. This document provides a more detailed discussion and instructions for using SAPHIRE 5.0 only when enough information on a specific topic is not provided by another available source

  4. Space Biology and Medicine. Volume 4; Health, Performance, and Safety of Space Crews

    Science.gov (United States)

    Dietlein, Lawrence F. (Editor); Pestov, Igor D. (Editor)

    2004-01-01

    Volume IV is devoted to examining the medical and associated organizational measures used to maintain the health of space crews and to support their performance before, during, and after space flight. These measures, collectively known as the medical flight support system, are important contributors to the safety and success of space flight. The contributions of space hardware and the spacecraft environment to flight safety and mission success are covered in previous volumes of the Space Biology and Medicine series. In Volume IV, we address means of improving the reliability of people who are required to function in the unfamiliar environment of space flight as well as the importance of those who support the crew. Please note that the extensive collaboration between Russian and American teams for this volume of work resulted in a timeframe of publication longer than originally anticipated. Therefore, new research or insights may have emerged since the authors composed their chapters and references. This volume includes a list of authors' names and addresses should readers seek specifics on new information. At least three groups of factors act to perturb human physiological homeostasis during space flight. All have significant influence on health, psychological, and emotional status, tolerance, and work capacity. The first and most important of these factors is weightlessness, the most specific and radical change in the ambient environment; it causes a variety of functional and structural changes in human physiology. The second group of factors precludes the constraints associated with living in the sealed, confined environment of spacecraft. Although these factors are not unique to space flight, the limitations they entail in terms of an uncomfortable environment can diminish the well-being and performance of crewmembers in space. The third group of factors includes the occupational and social factors associated with the difficult, critical nature of the

  5. Site Environmental Report for 2004. Volume 1, Environment, Health, and Safety Division

    Energy Technology Data Exchange (ETDEWEB)

    None

    2005-09-30

    Each year, Ernest Orlando Lawrence Berkeley National Laboratory prepares an integrated report on its environmental programs to satisfy the requirements of United States Department of Energy Order 231.1A, Environment, Safety, and Health Reporting.1 The Site Environmental Report for 2004 summarizes Berkeley Lab’s environmental management performance, presents environmental monitoring results, and describes significant programs for calendar year 2004. (Throughout this report, Ernest Orlando Lawrence Berkeley National Laboratory is referred to as “Berkeley Lab,” “the Laboratory,” “Lawrence Berkeley National Laboratory,” and “LBNL.”) The report is separated into two volumes. Volume I contains an overview of the Laboratory, the status of environmental programs, and summarized results from surveillance and monitoring activities. Volume II contains individual data results from these activities. This year, the Site Environmental Report was distributed by releasing it on the Web from the Berkeley Lab Environmental Services Group (ESG) home page, which is located at http://www.lbl.gov/ehs/esg/. Many of the documents cited in this report also are accessible from the ESG Web page. CD and printed copies of this Site Environmental Report are available upon request.

  6. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 1

    International Nuclear Information System (INIS)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains a discussion of the chemical safety improvements planned or already underway at DOE sites to correct facility or site-specific vulnerabilities. The main part of the report is a discussion of each of the programmatic deficiencies; a description of the tasks to be accomplished; the specific actions to be taken; and the organizational responsibilities for implementation

  7. Sensitivity and uncertainty analyses applied to criticality safety validation, methods development. Volume 1

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Hopper, C.M.; Childs, R.L.; Parks, C.V.

    1999-01-01

    This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the available S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently used by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The S/U methods that are presented in this volume are designed to provide a formal means of establishing the range (or area) of applicability for criticality safety data validation studies. The development of parameters that are analogous to the standard trending parameters forms the key to the technique. These parameters are the D parameters, which represent the differences by group of sensitivity profiles, and the ck parameters, which are the correlation coefficients for the calculational uncertainties between systems; each set of parameters gives information relative to the similarity between pairs of selected systems, e.g., a critical experiment and a specific real-world system (the application)

  8. Federal Logistics Information System (FLIS) Procedures Manual. Volume 3. Development and Maintenance of Item Logistics Data Tools.

    Science.gov (United States)

    1995-01-01

    Ammunition for cannon with illu- to fight fires minating projectile started by Ammunition for cannon with incen- explosion. diary projectile...OSNABURG Ozu Aya JEAN 146 DoD 4100.39-M Volume 3 APPENDIX 3-3-B SAMPLE OF NEW CONCEPT FIIG FIIG SAMPLE INC 00000 APPENDIX C Table 32 Palm Beach PLAIN

  9. Science Library of Test Items. Volume Eight. Mastery Testing Program. Series 3 & 4 Supplements to Introduction and Manual.

    Science.gov (United States)

    New South Wales Dept. of Education, Sydney (Australia).

    Continuing a series of short tests aimed at measuring student mastery of specific skills in the natural sciences, this supplementary volume includes teachers' notes, a users' guide and inspection copies of test items 27 to 50. Answer keys and test scoring statistics are provided. The items are designed for grades 7 through 10, and a list of the…

  10. Electrical safety code manual a plan language guide to national electrical code, OSHA and NFPA 70E

    CERN Document Server

    Keller, Kimberley

    2010-01-01

    Safety in any workplace is extremely important. In the case of the electrical industry, safety is critical and the codes and regulations which determine safe practices are both diverse and complicated. Employers, electricians, electrical system designers, inspectors, engineers and architects must comply with safety standards listed in the National Electrical Code, OSHA and NFPA 70E. Unfortunately, the publications which list these safety requirements are written in very technically advanced terms and the average person has an extremely difficult time understanding exactly what they need to

  11. Environmental, health, and safety issues of sodium-sulfur batteries for electric and hybrid vehicles. Volume 1, Cell and battery safety

    Energy Technology Data Exchange (ETDEWEB)

    Ohi, J M

    1992-09-01

    This report is the first of four volumes that identify and assess the environmental, health, and safety issues involved in using sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles that may affect the commercialization of Na/S batteries. This and the other reports on recycling, shipping, and vehicle safety are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD&D) program for Na/S battery technology. The reports review the status of Na/S battery RD&D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers cell design and engineering as the basis of safety for Na/S batteries and describes and assesses the potential chemical, electrical, and thermal hazards and risks of Na/S cells and batteries as well as the RD&D performed, under way, or to address these hazards and risks. The report is based on a review of the literature and on discussions with experts at DOE, national laboratories and agencies, universities, and private industry. Subsequent volumes will address environmental, health, and safety issues involved in shipping cells and batteries, using batteries to propel electric vehicles, and recycling and disposing of spent batteries. The remainder of this volume is divided into two major sections on safety at the cell and battery levels. The section on Na/S cells describes major component and potential failure modes, design, life testing and failure testing, thermal cycling, and the safety status of Na/S cells. The section on batteries describes battery design, testing, and safety status. Additional EH&S information on Na/S batteries is provided in the appendices.

  12. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

  13. Federal Logistics Information System (FLIS) Procedures Manual. Volume 8. Document Identifier Code Input/Output Formats (Fixed Length)

    Science.gov (United States)

    1994-07-01

    REQUIRED MIX OF SEGMENTS OR INDIVIDUAL DATA ELEMENTS TO BE EXTRACTED. IN SEGMENT R ON AN INTERROGATION TRANSACTION (LTI), DATA RECORD NUMBER (DRN 0950) ONLY...and zation and Marketing input DICs. insert the Continuation Indicator Code (DRN 8555) in position 80 of this record. Maximum of OF The assigned NSN...for Procurement KFR, File Data Minus Security Classified Characteristics Data KFC 8.5-2 DoD 4100.39-M Volume 8 CHAPTER 5 ALPHABETIC INDEX OF DIC

  14. Annual Conference on Manual Control (20th) Held in California on June 12 - 14, 1984. Volume 1

    Science.gov (United States)

    1984-06-14

    Semisupine Pilot", Avimtion Space and Environmental Medicine , vol. 49, No. 1, pp. 304-317, January, 1978. [5] Levison, W.H., "Model For Human Controller...Performance in Vibration Environment", Aviation Space and Environmental Medicine , vol. 49, No. 1, pp. 321-327, January, 1978. [6] Repperger, D. W...Schwartz, Anthony C. Stein, Raymond E. Magdaleno, and Jeffrey R. Hogue, Effects of Alcohol and Marihuana on Driver Control Behavior. Volume I: Laboratory

  15. Associations between Family Adversity and Brain Volume in Adolescence: Manual vs. Automated Brain Segmentation Yields Different Results

    OpenAIRE

    Lyden, Hannah; Gimbel, Sarah I.; Del Piero, Larissa; Tsai, A. Bryna; Sachs, Matthew E.; Kaplan, Jonas T.; Margolin, Gayla; Saxbe, Darby

    2016-01-01

    Associations between brain structure and early adversity have been inconsistent in the literature. These inconsistencies may be partially due to methodological differences. Different methods of brain segmentation may produce different results, obscuring the relationship between early adversity and brain volume. Moreover, adolescence is a time of significant brain growth and certain brain areas have distinct rates of development, which may compromise the accuracy of automated segmentation appr...

  16. Databases on safety issues for WWER and RBMK reactors. Users' manual. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-04-01

    At the beginning of the IAEA Extrabudgetary Programme on the safety of WWER reactors a great number of findings and recommendations (safety items) were collected as a result of design review and safety review missions of the WWER-440/230 type reactors. On the basis of these findings a technical database containing more than 1300 records was established to support the consolidation of the information obtained and to help in identification of safety issues. After the scope of the WWER extrabudgetary programme was extended similar data sets were prepared for the WWER-440/213, WWER-1000 and RBMK nuclear power plants. This publication describes the structure of the databases on safety issues of WWER and RBMK NPPs, the information sources used in the databases and interrogation capabilities for users to obtain the necessary information. 14 refs, 9 figs, 5 tabs

  17. Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

  18. Surface Environmental Surveillance Project: Locations Manual Volume 1 – Air and Water Volume 2 – Farm Products, Soil & Vegetation, and Wildlife

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, Brad G.; Patton, Gregory W.; Stegen, Amanda; Poston, Ted M.

    2009-01-01

    This report describes all environmental monitoring locations associated with the Surface Environmental Surveillance Project. Environmental surveillance of the Hanford site and surrounding areas is conducted by the Pacific Northwest National Laboratory (PNNL) for the U.S. Department of Energy (DOE). Sampling is conducted to evaluate levels of radioactive and nonradioactive pollutants in the Hanford environs, as required in DOE Order 450.1, Environmental Protection Program, and DOE Order 5400.5, Radiation Protection of the Public and the Environment. The environmental surveillance sampling design is described in the Hanford Site Environmental Monitoring Plan, United States Department of Energy, Richland Operation Office (DOE/RL-91-50). This document contains the locations of sites used to collect samples for the Surface Environmental Surveillance Project (SESP). Each section includes directions, maps, and pictures of the locations. A general knowledge of roads and highways on and around the Hanford Site is necessary to successfully use this manual. Supplemental information (Maps, Gazetteer, etc.) may be necessary if user is unfamiliar with local routes. The SESP is a multimedia environmental surveillance effort to measure the concentrations of radionuclides and chemicals in environmental media to demonstrate compliance with applicable environmental quality standards and public exposure limits, and assessing environmental impacts. Project personnel annually collect selected samples of ambient air, surface water, agricultural products, fish, wildlife, and sediments. Soil and vegetation samples are collected approximately every 5 years. Analytical capabilities include the measurement of radionuclides at very low environmental concentrations and, in selected media, nonradiological chemicals including metals, anions, volatile organic compounds, and total organic carbon.

  19. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  20. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  1. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 2

    International Nuclear Information System (INIS)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. To address the facility-specific and site-specific vulnerabilities, responsible DOE and site-contractor line organizations have developed initial site response plans. These plans, presented as Volume 2 of this Management Response Plan, describe the actions needed to mitigate or eliminate the facility- and site-specific vulnerabilities identified by the CSV Working Group field verification teams. Initial site response plans are described for: Brookhaven National Lab., Hanford Site, Idaho National Engineering Lab., Lawrence Livermore National Lab., Los Alamos National Lab., Oak Ridge Reservation, Rocky Flats Plant, Sandia National Laboratories, and Savannah River Site

  2. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  3. TIBER II/ETR final design report: Volume 3, 5.0 Radiation safety and environment; 6.0 Physics and technology R and D needs

    International Nuclear Information System (INIS)

    Lee, J.D.

    1987-09-01

    This paper discusses the design of the TIBER II Tokamak. This particular volume discusses: safety and environmental requirements and design targets; accident analyses; personnel safety and maintenance exposure; effluent control; waste management and decommissioning; safety considerations in building design; and safety and environmental conclusions and recommendations

  4. Fire Protection Program Manual

    Energy Technology Data Exchange (ETDEWEB)

    Sharry, J A

    2012-05-18

    This manual documents the Lawrence Livermore National Laboratory (LLNL) Fire Protection Program. Department of Energy (DOE) Orders 420.1B, Facility Safety, requires LLNL to have a comprehensive and effective fire protection program that protects LLNL personnel and property, the public and the environment. The manual provides LLNL and its facilities with general information and guidance for meeting DOE 420.1B requirements. The recommended readers for this manual are: fire protection officers, fire protection engineers, fire fighters, facility managers, directorage assurance managers, facility coordinators, and ES and H team members.

  5. Knowledge of Health Volunteers in the Damavand District on Food Safety: A Study Based on the World Health Organization Manual

    Directory of Open Access Journals (Sweden)

    M Ghfari

    2016-09-01

    Full Text Available Introduction: Food borne illnesses are usually infectious or toxic in nature and caused by bacteria, viruses, parasites or chemical substances entering the body through contaminated food or water. Unsafe food causes more than 200 diseases - ranging from diarrhea to cancers. Food safety is a public health priority. The aim of present study was determine level of health communicators' knowledge about food safety in Damavand city that was performed according to the World Health Organization guide. Methods: This study is a descriptive analysis of the target a group of health Volunteers in Damavand depended to Shahid Beheshti University of Medical Sciences. 109 persons were enrolled with s awareness, scarification and consent of the census. Tools for data collection was a questionnaire consisting of 13 questions on demographic characteristics and 24 questions about food safety according to WHO’s guide. Data was analyzed with SPSS-18 software and statistical analysis includes one-way ANOVA, t-test and spearman correlation. Results: The mean age of participant was 44.75 ± 9.98 years. Average score of awareness of food safety was 35.87 ±6.22 and for awareness was 77 (71%. No significant relationships was observed between of awareness and marital status, age, education, occupation and education of wife was not significant relationships (p>0/05. Conclusion: In some safety food items the level of knowledge in some safety food items was good, in other one, such as food storage, transmission of microbes, cooking temperature for meat there was low awareness. In this respect, information, education programs to raise awareness of the health status for health volunteers is recommended

  6. Safety assessment methodologies for near surface disposal facilities. Results of a co-ordinated research project (ISAM). Volume 1: Review and enhancement of safety assessment approaches and tools. Volume 2: Test cases

    International Nuclear Information System (INIS)

    2004-07-01

    the Safety Guide on 'Safety Assessment for Near Surface Disposal of Radioactive Waste' (Safety Standards Series No. WS-G- 1.1). The report of this CRP is presented in two volumes; Volume 1 contains a summary and a complete description of the ISAM project methodology and Volume 2 presents the application of the methodology to three hypothetical test cases

  7. Guidelines for the verification and validation of expert system software and conventional software. Volume 7, User's manual: Final report

    International Nuclear Information System (INIS)

    Miller, L.A.; Hayes, J.E.; Mirsky, S.M.

    1995-05-01

    Reliable software is required for nuclear power industry applications. Verification and validation techniques applied during the software development process can help eliminate errors that could inhibit the proper operation of digital systems and cause availability and safety problems. Most of the techniques described in this report are valid for conventional software systems as well as for expert systems. The project resulted in a set of 16 V ampersand V guideline packages and 11 sets of procedures based on the class, development phase, and system component being tested. These guideline packages and procedures help a utility define the level of V ampersand V, which involves evaluating the complexity and type of software component along with the consequences of failure. In all, the project identified 153 V ampersand V techniques for conventional software systems and demonstrated their application to all aspects of expert systems except for the knowledge base, which requires specially developed tools. Each of these conventional techniques covers anywhere from 2-52 total types of conventional software defects, and each defect is covered by 21-50 V ampersand V techniques. The project also identified automated tools to Support V ampersand V activities

  8. SMACS: a system of computer programs for probabilistic seismic analysis of structures and subsystems. Volume I. User's manual

    International Nuclear Information System (INIS)

    Maslenikov, O.R.; Johnson, J.J.; Tiong, L.W.; Mraz, M.J.; Bumpus, S.; Gerhard, M.A.

    1985-03-01

    The SMACS (Seismic Methodology Analysis Chain with Statistics) system of computer programs, one of the major computational tools of the Seismic Safety Margins Research Program (SSMRP), links the seismic input with the calculation of soil-structure interaction, major structure response, and subsystem response. The seismic input is defined by ensembles of acceleration time histories in three orthogonal directions. Soil-structure interaction and detailed structural response are then determined simultaneously, using the substructure approach to SSI as implemented in the CLASSI family of computer programs. The modus operandi of SMACS is to perform repeated deterministic analyses, each analysis simulating an earthquake occurrence. Parameter values for each simulation are sampled from assumed probability distributions according to a Latin hypercube experimental design. The user may specify values of the coefficients of variation (COV) for the distributions of the input variables. At the heart of the SMACS system is the computer program SMAX, which performs the repeated SSI response calculations for major structure and subsystem response. This report describes SMAX and the pre- and post-processor codes, used in conjunction with it, that comprise the SMACS system

  9. Estudo experimental sobre a eficiência e segurança da manobra de hiperinsuflação manual como técnica de remoção de secreção Experimental study on the efficiency and safety of the manual hyperinflation maneuver as a secretion clearance technique

    Directory of Open Access Journals (Sweden)

    Tatiana de Arruda Ortiz

    2013-04-01

    Full Text Available OBJETIVO: Avaliar, em um modelo pulmonar simulando um paciente sob ventilação mecânica, a eficiência e a segurança da manobra de hiperinsuflação manual (HM com o intuito de remover secreção pulmonar. MÉTODOS: Oito fisioterapeutas utilizaram um ressuscitador manual autoinflável para realizar HM com o objetivo de remover secreções, em duas condições: conforme rotineiramente aplicada durante sua prática clínica, e após receberem instruções verbais baseadas em recomendações de especialistas. Três cenários clínicos foram simulados: função pulmonar normal, doença pulmonar restritiva e doença pulmonar obstrutiva. RESULTADOS: Antes da instrução, o uso de duas compressões sequenciais do ressuscitador era comum, e a pressão proximal (Pprox foi mais alta em relação à obtida após a instrução. Entretanto, a pressão alveolar (Palv nunca excedeu 42,5 cmH2O (mediana, 16,1; intervalo interquartil [IQ], 11,7-24,5, mesmo com valores de Pprox de até 96,6 cmH2O (mediana, 36,7; IQ, 22,9-49,4. O volume corrente (VC gerado foi relativamente pequeno (mediana, 640 mL; IQ, 505-735 e o pico de fluxo inspiratório (PFI geralmente excedeu o pico de fluxo expiratório (PFE: 1,37 L/s (IQ, 0,99-1,90 e 1,01 L/s (IQ, 0,55-1,28, respectivamente. Uma relação PFI/PFE OBJECTIVE: To evaluate, in a lung model simulating a mechanically ventilated patient, the efficiency and safety of the manual hyperinflation (MH maneuver as a means of removing pulmonary secretions. METHODS: Eight respiratory therapists (RTs were asked to use a self-inflating manual resuscitator on a lung model to perform MH as if to remove secretions, under two conditions: as routinely applied during their clinical practice; and after receiving verbal instructions based on expert recommendations. In both conditions, three clinical scenarios were simulated: normal lung function, restrictive lung disease, and obstructive lung disease. RESULTS: Before instruction, it was common

  10. Users manual for Aerospace Nuclear Safety Program six-degree-of-freedom reentry simulation (TMAGRA6C)

    International Nuclear Information System (INIS)

    Sharbaugh, R.C.

    1990-02-01

    This report documents the updated six-degree-of-freedom reentry simulation TMAGRA6C used in the Aerospace Nuclear Safety Program, ANSP. The simulation provides for the inclusion of the effects of ablation on the aerodynamic stability and drag of reentry bodies, specifically the General Purpose Heat Source, GPHS. The existing six-degree-of-freedom reentry body simulations (TMAGRA6A and TMAGRA6B) used in the JHU/APL Nuclear Safety Program do not include aerodynamic effects resulting from geometric changes to the configuration due to ablation from reentry flights. A wind tunnel test was conducted in 1989 to obtain the effects of ablation on the hypersonic aerodynamics of the GPHS module. The analyzed data were used to form data sets which are included herein in tabular form. These are used as incremental aerodynamic inputs in the new TMAGRA6C six-degree-of-freedom reentry simulation. 20 refs., 13 figs., 2 tabs

  11. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0: Integrated Reliability and Risk Analysis System (IRRAS) reference manual. Volume 2

    International Nuclear Information System (INIS)

    Russell, K.D.; Kvarfordt, K.J.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

    1994-07-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the use the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification to report generation. Version 1.0 of the IRRAS program was released in February of 1987. Since then, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 5.0 and is the subject of this Reference Manual. Version 5.0 of IRRAS provides the same capabilities as earlier versions and ads the ability to perform location transformations, seismic analysis, and provides enhancements to the user interface as well as improved algorithm performance. Additionally, version 5.0 contains new alphanumeric fault tree and event used for event tree rules, recovery rules, and end state partitioning

  12. The safety of dipyridamole in patients undergoing myocardial perfusion scintigraphy prior to lung volume reduction surgery

    International Nuclear Information System (INIS)

    Roman, M.R.; Angelides, S.; Parker, M.K.; Silva, I. da; Freeman, A.P.

    2001-01-01

    Patients with end-stage chronic obstructive pulmonary disease (COPD) undergoing lung volume reduction surgery (LVRS) are at high risk of peri-operative cardiac complications, and myocardial perfusion scintigraphy (MPS) is commonly used for risk stratification. This study prospectively assessed the safety of dipyridamole in these patients and compared the incidence of side-effects (particularly dyspnoea) with that in patients undergoing dipyridamole MPS prior to elective non-cardiothoracic surgery. Fifty patients were enrolled: 25 in the LVRS cohort (13 males, 12 females), with a mean age of 65 years and a mean FEV 1 of 0.79 l, and 25 (with no history of asthma or COPD) in the control cohort (14 males, 11 females), with a mean age of 66 years. Fourteen patients (56%) in each group developed side-effects. Dyspnoea was reported by five patients (20%) in the LVRS and two patients (8%) in the control cohort (P=NS). One patient in each cohort developed severe hypotension and bradycardia. Eight (32%) other patients developed minor side-effects in the LVRS cohort compared with 11 (44%) in the control group. All side-effects responded promptly to intravenous aminophylline. In summary, there was a statistically non-significant increase in the incidence of dyspnoea in patients with end-stage COPD and all side-effects responded to aminophylline. Thus, dipyridamole can be used safely in these patients. (orig.)

  13. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    International Nuclear Information System (INIS)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  14. Safety of High Speed Ground Transportation Systems : Analytical Methodology for Safety Validation of Computer Controlled Subsystems : Volume 2. Development of a Safety Validation Methodology

    Science.gov (United States)

    1995-01-01

    This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety cortical functions in high-speed rail or magnetic levitation ...

  15. Life Skills Coach Training Manual.

    Science.gov (United States)

    Saskatchewan NewStart, Inc., Prince Albert.

    Ways of helping coaches to counsel unemployed adults in the solving of their personal problems are explored in this manual. Originally printed as two separate volumes, this reprinting of the study has bound the two together. Volume I involves a general discussion of life's problems and of the need to solve them. This volume contains four parts.…

  16. Environmental contamination by cyclophosphamide preparation: Comparison of conventional manual production in biological safety cabinet and robot-assisted production by APOTECAchemo.

    Science.gov (United States)

    Schierl, Rudolf; Masini, Carla; Groeneveld, Svenja; Fischer, Elke; Böhlandt, Antje; Rosini, Valeria; Paolucci, Demis

    2016-02-01

    The aim of this study was to compare environmental contamination of cyclophosphamide (CP) during 1 week of drug compounding by conventional manual procedure in a biological safety cabinet (BSC) with laminar airflow and a new robotic drug preparation system (APOTECAchemo). During four consecutive days, similar numbers of infusion bags with cyclophosphamide were prepared with both techniques in a cross-over design. Wipe samples (49 for BSC, 50 for APOTECAchemo) were taken at several locations (gloves, infusion bags, trays, BSC-benches, floor) in the pharmacy and analyzed for CP concentrations by GC-MSMS (LOD 0.2 ng/sample). The detection rate was 70% in the BSC versus 15% in APOTECAchemo. During manual preparation of admixtures using BSC contamination with CP was below 0.001 ng/cm(2) at most locations, but significant on gloves (0.0004-0.0967 ng/cm(2)) and the majority (70%) of infusion bags (preparation by APOTECAchemo, gloves (1 of 8: 0.0007 ng/cm(2)) and infusion bags (3 of 20: 0.0005, 0.0019, 0.0094 ng/cm(2)) were considerably less contaminated. Residual contamination was found on the surfaces under the dosing device in the compounding area (0.0293-0.1603 ng/cm(2)) inside the robotic system. Compared to outcomes of other studies, our results underline good manufacturing procedures in this pharmacy with low contamination for both techniques (BSC and APOTECAchemo). Comparison of both preparation procedures validated that contamination of infusion bags was much lower by using the robotic system. © The Author(s) 2014.

  17. Manual on industrial radiography

    International Nuclear Information System (INIS)

    1981-08-01

    This manual is intended as a source of educational material to personnel seeking certification as industrial radiographers, and as a guide and reference text for educational organizations that are providng courses in industrial radiography. It covers the basic principles of x-ray and gamma radiation, radiation safety, films and film processing, welding, casting and forging, aircraft structures and components, radiographic techniques, and records

  18. Methodology for assessing the safety of Hydrogen Systems: HyRAM 1.1 technical reference manual

    Energy Technology Data Exchange (ETDEWEB)

    Groth, Katrina; Hecht, Ethan; Reynolds, John Thomas; Blaylock, Myra L.; Erin E. Carrier

    2017-03-01

    The HyRAM software toolkit provides a basis for conducting quantitative risk assessment and consequence modeling for hydrogen infrastructure and transportation systems. HyRAM is designed to facilitate the use of state-of-the-art science and engineering models to conduct robust, repeatable assessments of hydrogen safety, hazards, and risk. HyRAM is envisioned as a unifying platform combining validated, analytical models of hydrogen behavior, a stan- dardized, transparent QRA approach, and engineering models and generic data for hydrogen installations. HyRAM is being developed at Sandia National Laboratories for the U. S. De- partment of Energy to increase access to technical data about hydrogen safety and to enable the use of that data to support development and revision of national and international codes and standards. This document provides a description of the methodology and models contained in the HyRAM version 1.1. HyRAM 1.1 includes generic probabilities for hydrogen equipment fail- ures, probabilistic models for the impact of heat flux on humans and structures, and computa- tionally and experimentally validated analytical and first order models of hydrogen release and flame physics. HyRAM 1.1 integrates deterministic and probabilistic models for quantifying accident scenarios, predicting physical effects, and characterizing hydrogen hazards (thermal effects from jet fires, overpressure effects from deflagrations), and assessing impact on people and structures. HyRAM is a prototype software in active development and thus the models and data may change. This report will be updated at appropriate developmental intervals.

  19. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  20. Transportation security personnel training manual

    International Nuclear Information System (INIS)

    1978-11-01

    This volume is the instructor's manual for the training of SNM guards. Covered are: self-defense, arrest authority, civil liability, report writing, stress, tactics, and situational training scenarios

  1. Transportation security personnel training manual

    International Nuclear Information System (INIS)

    1978-11-01

    Objective of this manual is to train security personnel to protect special nuclear materials and nuclear facilities against theft and sabotage as required by 10 CFR Part 73. This volume contains the introduction and rationale

  2. 2007 motor vehicle occupant safety survey. Volume 3, air bags report

    Science.gov (United States)

    2008-11-01

    The 2007 Motor Vehicle Occupant Safety Survey was the sixth in a series of periodic national telephone surveys on occupant : protection issues conducted for the National Highway Traffic Safety Administration (NHTSA). Data collection was conducted : b...

  3. Jefferson Proving Ground, South of the Firing Line Health and Safety Plan, Volume 4

    National Research Council Canada - National Science Library

    1992-01-01

    .... The purpose of this Site Health and Safety Plan (SHSP) is to assign SECD personnel health and safety responsibilities, to prescribe mandatory operating procedures, and to establish personal-protective-equipment (PPE...

  4. Volume 8 No. 4 2008 December 2008 427 THE USE OF SAFETY ...

    African Journals Online (AJOL)

    the knowledge and use of safety devices and methods attached to the proper application of the ... farmers aware of the high risk due to inadequate education, training and safety devices in ..... delivery in Southwestern Nigeria. Ph. D Thesis.

  5. Manual Therapy

    OpenAIRE

    Hakgüder, Aral; Kokino, Siranuş

    2002-01-01

    Manual therapy has been used in the treatment of pain and dysfunction of spinal and peripheral joints for more than a hundred years. Manual medicine includes manipulation, mobilization, and postisometric relaxation techniques. The aim of manual therapy is to enhance restricted movement caused by blockage of joints keeping postural balance, restore function and maintain optimal body mechanics. Anatomic, biomechanical, and neurophysiological evaluations of the leucomotor system is essential for...

  6. Dose-Volume Histogram Analysis of the Safety of Proton Beam Therapy for Unresectable Hepatocellular Carcinoma

    International Nuclear Information System (INIS)

    Kawashima, Mitsuhiko; Kohno, Ryosuke; Nakachi, Kohei; Nishio, Teiji; Mitsunaga, Shuichi; Ikeda, Masafumi; Konishi, Masaru; Takahashi, Shinichiro; Gotohda, Naoto; Arahira, Satoko; Zenda, Sadamoto; Ogino, Takashi; Kinoshita, Taira

    2011-01-01

    Purpose: To evaluate the safety and efficacy of radiotherapy using proton beam (PRT) for unresectable hepatocellular carcinoma. Methods and Materials: Sixty consecutive patients who underwent PRT between May 1999 and July 2007 were analyzed. There were 42 males and 18 females, with a median age of 70 years (48-92 years). All but 1 patient had a single lesion with a median diameter of 45 mm (20-100 mm). Total PRT dose/fractionation was 76-cobalt Gray equivalent (CGE)/20 fractions in 46 patients, 65 CGE/26 fractions in 11 patients, and 60 CGE/10 fractions in 3 patients. The risk of developing proton-induced hepatic insufficiency (PHI) was estimated using dose-volume histograms and an indocyanine-green retention rate at 15 minutes (ICG R15). Results: None of the 20 patients with ICG R15 of less than 20% developed PHI, whereas 6 of 8 patients with ICG R15 values of 50% or higher developed PHI. Among 32 patients whose ICG R15 ranged from 20% to 49.9%, PHI was observed only in patients who had received 30 CGE (V30) to more than 25% of the noncancerous parts of the liver (n = 5) Local progression-free and overall survival rates at 3 years were 90% (95% confidence interval [CI], 80-99%) and 56% (95% CI, 43-69%), respectively. A gastrointestinal toxicity of Grade ≥2 was observed in 3 patients. Conclusions: ICG R15 and V30 are recommended as useful predictors for the risk of developing PHI, which should be incorporated into multidisciplinary treatment plans for patients with this disease.

  7. Between demarcation and discretion: The medical-administrative boundary as a locus of safety in high-volume organisational routines.

    Science.gov (United States)

    Grant, Suzanne; Guthrie, Bruce

    2018-04-01

    Patient safety is an increasing concern for health systems internationally. The majority of administrative work in UK general practice takes place in the context of organisational routines such as repeat prescribing and test results handling, where high workloads and increased clinician dependency on administrative staff have been identified as an emerging safety issue. Despite this trend, most research to date has focused on the redistribution of the clinical workload between doctors, nurses and allied health professionals within individual care settings. Drawing on Strauss's negotiated order perspective, we examine ethnographically the achievement of safety across the medical-administrative boundary in key high-volume routines in UK general practice. We focus on two main issues. First, GPs engaged in strategies of demarcation by defining receptionist work as routine, unspecialised and dependent upon GP clinical knowledge and oversight as the safety net to deal with complexity and risk. Receptionists consented to this 'social closure' when describing their role, thus reinforcing the underlying inter-occupational relationship of medical domination. Second, in everyday practice, GPs and receptionists engaged in informal boundary-blurring to safely accommodate the complexity of everyday high-volume routine work. This comprised additional informal discretionary spaces for receptionist decision-making and action that went beyond the routine safety work formally assigned to them. New restratified intra-occupational hierarchies were also being created between receptionists based on the complexity of the safety work that they were authorised to do at practice level, with specialised roles constituting a new form of administrative 'professional project'. The article advances negotiated order theory by providing an in-depth examination of the ways in which medical-administrative boundary-making and boundary-blurring constitute distinct modes of safety in high-volume

  8. Sandia SWiFT Wind Turbine Manual.

    Energy Technology Data Exchange (ETDEWEB)

    White, Jonathan; LeBlanc, Bruce Philip; Berg, Jonathan Charles; Bryant, Joshua; Johnson, Wesley D.; Paquette, Joshua

    2016-01-01

    The Scaled Wind Farm Technology (SWiFT) facility, operated by Sandia National Laboratories for the U.S. Department of Energy's Wind and Water Power Program, is a wind energy research site with multiple wind turbines scaled for the experimental study of wake dynamics, advanced rotor development, turbine control, and advanced sensing for production-scale wind farms. The SWiFT site currently includes three variable-speed, pitch-regulated, three-bladed wind turbines. The six volumes of this manual provide a detailed description of the SWiFT wind turbines, including their operation and user interfaces, electrical and mechanical systems, assembly and commissioning procedures, and safety systems. Further dissemination only as authorized to U.S. Government agencies and their contractors; other requests shall be approved by the originating facility or higher DOE programmatic authority. 111 UNCLASSIFIED UNLIMITED RELEASE Sandia SWiFT Wind Turbine Manual (SAND2016-0746 ) approved by: Department Manager SWiFT Site Lead Dave Minster (6121) Date Jonathan White (6121) Date SWiFT Site Supervisor Dave Mitchell (6121) Date Note: Document revision logs are found after the title page of each volume of this manual. iv

  9. Environmental, health, and safety issues of sodium-sulfur batteries for electric and hybrid vehicles. Volume 4, In-vehicle safety

    Energy Technology Data Exchange (ETDEWEB)

    Mark, J.

    1992-11-01

    This report is the last of four volumes that identify and assess the environmental, health, and safety issues that may affect the commercial-scale use of sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles. The reports are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD&D) program for Na/S battery technology. The reports review the status of Na/S battery RD&D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers the in-vehicle safety issues of electric vehicles powered by Na/S batteries. The report is based on a review of the literature and on discussions with experts at DOE, national laboratories and agencies, and private industry. It has three major goals: (1) to identify the unique hazards associated with electric vehicle (EV) use; (2) to describe the existing standards, regulations, and guidelines that are or could be applicable to these hazards; and (3) to discuss the adequacy of the existing requirements in addressing the safety concerns of EVs.

  10. Design data and safety features of commercial nuclear power plants including cumulative index for Volumes I--VI

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1977-01-01

    Design data, safety features, and site characteristics are summarized for 12 nuclear power units in 6 power stations in the United States. Six pages of data are presented for each station, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. In addition, an aerial perspective is presented for each plant. This volume covers plants with docket numbers 50-553 through 50-569 (Phipps Bend, Black Fox, Yellow Creek, and NEP) and two earlier plants not previously reported--Hope Creek (50-354, 50-355) and WPPSS 1 and 4 (50-460, 50-513). Indexes for this volume and the five earlier volumes are presented in three forms--by docket number, by plant name, and by participating utility

  11. Structures manual

    Science.gov (United States)

    2001-01-01

    This manual was written as a guide for use by design personnel in the Vermont Agency of Transportation Structures Section. This manual covers the design responsibilities of the Section. It does not cover other functions that are a part of the Structu...

  12. Quality Manual

    Science.gov (United States)

    Koch, Michael

    The quality manual is the “heart” of every management system related to quality. Quality assurance in analytical laboratories is most frequently linked with ISO/IEC 17025, which lists the standard requirements for a quality manual. In this chapter examples are used to demonstrate, how these requirements can be met. But, certainly, there are many other ways to do this.

  13. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  14. Nuclear Safety. Technical Progress Journal, October--December 1992: Volume 33, No. 4

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  15. Nuclear Safety. Technical progress journal, January--March 1994: Volume 35, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1994-01-01

    This is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  16. Nuclear Safety. Technical progress journal, April--June 1992: Volume 33, No.2

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  17. Nuclear Safety. Technical Progress Journal, January--March 1993: Volume 34, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  18. Nuclear Safety. Technical progress journal, April--June 1996: Volume 37, No. 2

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, M D [ed.

    1996-01-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  19. Nuclear Safety. Technical Progress Journal, January--March 1992: Volume 33, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  20. Nuclear safety. Technical progress journal, January--March 1996: Volume 37, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, M D [ed.

    1996-01-01

    Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but is also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Individual articles are indexed separately for the data base.

  1. Radiological control manual. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kloepping, R.

    1996-05-01

    This Lawrence Berkeley National Laboratory Radiological Control Manual (LBNL RCM) has been prepared to provide guidance for site-specific additions, supplements and interpretation of the DOE Radiological Control Manual. The guidance provided in this manual is one methodology to implement the requirements given in Title 10 Code of Federal Regulations Part 835 (10 CFR 835) and the DOE Radiological Control Manual. Information given in this manual is also intended to provide demonstration of compliance to specific requirements in 10 CFR 835. The LBNL RCM (Publication 3113) and LBNL Health and Safety Manual Publication-3000 form the technical basis for the LBNL RPP and will be revised as necessary to ensure that current requirements from Rules and Orders are represented. The LBNL RCM will form the standard for excellence in the implementation of the LBNL RPP.

  2. Radiological control manual. Revision 1

    International Nuclear Information System (INIS)

    Kloepping, R.

    1996-05-01

    This Lawrence Berkeley National Laboratory Radiological Control Manual (LBNL RCM) has been prepared to provide guidance for site-specific additions, supplements and interpretation of the DOE Radiological Control Manual. The guidance provided in this manual is one methodology to implement the requirements given in Title 10 Code of Federal Regulations Part 835 (10 CFR 835) and the DOE Radiological Control Manual. Information given in this manual is also intended to provide demonstration of compliance to specific requirements in 10 CFR 835. The LBNL RCM (Publication 3113) and LBNL Health and Safety Manual Publication-3000 form the technical basis for the LBNL RPP and will be revised as necessary to ensure that current requirements from Rules and Orders are represented. The LBNL RCM will form the standard for excellence in the implementation of the LBNL RPP

  3. Westinghouse Reference Safety Analysis Report, RESAR-414. License application, preliminary safety analysis report (RESAR-414) volume 1

    International Nuclear Information System (INIS)

    1976-01-01

    Westinghouse's standardized four-loop, single unit NSSS for a pressurized water reactor is described including the core, coolant system, ECCS, emergency boration, chemical and volume control, RHR system, boron recycle, fuel handling, spent fuel pool and associated instrumentation and controls. This reactor is applicable to a plant with a core power level of 3800 MW(t) and 1295 MW(e). The reactor is controlled by temperature coefficients of reactivity; control rod motion, and by a soluble neutron absorber-boric acid

  4. 1987 Oak Ridge model conference: Proceedings: Volume 3, Health and safety

    International Nuclear Information System (INIS)

    1987-01-01

    See the abstract for Volume I for general information on the conference. Topics discussed in Volume III include the use of models in handling hazardous materials, communication at waste sites, asbestos, regulatory decisions, emergency planning, training programs, occupational hazards, and protection of subcontractors

  5. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  6. Cardiac MRI in children and adolescents who have undergone surgical repair of right-sided congenital heart disease. Automated left ventricular volumes and function analysis and effects of different manual adjustments

    Energy Technology Data Exchange (ETDEWEB)

    Rompel, O.; Janka, R.; May, M.S.; Lell, M.M.; Uder, M.; Hammon, M. [University Hospital Erlangen (Germany). Dept. of Radiology; Gloeckler, M.; Dittrich, S. [University Hospital Erlangen (Germany). Dept. of Pediatric Cardiology; Cesnjevar, R. [University Hospital Erlangen (Germany). Dept. of Pediatric Cardiac Surgery

    2015-12-15

    To evaluate automated segmentation and the effects of different manual adjustments regarding left ventricular parameter quantification in cardiac magnetic resonance (MR) data on children and adolescents who have undergone surgical repair of right-sided congenital heart disease (CHD). Dedicated software (syngo.via, Siemens AG) was used to automatically segment and/or manually adjust the end-diastolic volume (EDV), end-systolic volume (ESV), stroke volume (SV), myocardial mass (MM) and ejection fraction (EF) before/after manual apex/base adjustment (ADJ-step 1) and after manual apex/base/myocardial contour adjustment (ADJ-step 2; reference standard). MR data of 40 patients (13.1 ± 3.1y, 4-17y) with repaired CHD with decreased pulmonary blood flow (CHD-DPBF) were evaluated. Intra- and inter-rater reliability was determined for 10 randomly selected patients. The software correctly detected the left ventricle in 38/40 (95 %) patients. EDV after automated segmentation: 119.1 ± 44.0ml; after ADJ-step 1: 115.8 ± 39.5 ml; after ADJ-step 2: 116.2 ± 39.4 ml. The corresponding results for ESV were 52.0 ± 18.5/49.6 ± 16.9/49.7 ± 16.4 ml; for SV 67.1 ± 28.5/66.2 ± 25.4/66.5 ± 25.5 ml; for EF 55.5 ± 7.3/56.7 ± 6.6/56.7 ± 6.3%; for MM 83.7 ± 35.9/76.2 ± 28.3/74.6 ± 27.2 g. Significant differences were found for ESV/MM/EF comparing the automated segmentation results with these after ADJ-step 1 and ADJ-step 2. No significant differences were found when comparing all results of ADJ-step 1 and ADJ-step 2 or when comparing EDV/SV results. Intra- and inter-rater reliability was excellent. The mean time effort was 63.4 ± 6.9 s for the automated segmentation, 74.2 ± 8.9 s for ADJ-step 1 and 269.5 ± 39.4 s for ADJ-step 2. Automated left ventricular volumes and function analysis in children and adolescents with surgically treated CHD proved to be feasible with excellent intra- and inter-rater reliability. Automated segmentation with manual apex/base adjustment provided

  7. Cardiac MRI in children and adolescents who have undergone surgical repair of right-sided congenital heart disease. Automated left ventricular volumes and function analysis and effects of different manual adjustments

    International Nuclear Information System (INIS)

    Rompel, O.; Janka, R.; May, M.S.; Lell, M.M.; Uder, M.; Hammon, M.; Gloeckler, M.; Dittrich, S.; Cesnjevar, R.

    2015-01-01

    To evaluate automated segmentation and the effects of different manual adjustments regarding left ventricular parameter quantification in cardiac magnetic resonance (MR) data on children and adolescents who have undergone surgical repair of right-sided congenital heart disease (CHD). Dedicated software (syngo.via, Siemens AG) was used to automatically segment and/or manually adjust the end-diastolic volume (EDV), end-systolic volume (ESV), stroke volume (SV), myocardial mass (MM) and ejection fraction (EF) before/after manual apex/base adjustment (ADJ-step 1) and after manual apex/base/myocardial contour adjustment (ADJ-step 2; reference standard). MR data of 40 patients (13.1 ± 3.1y, 4-17y) with repaired CHD with decreased pulmonary blood flow (CHD-DPBF) were evaluated. Intra- and inter-rater reliability was determined for 10 randomly selected patients. The software correctly detected the left ventricle in 38/40 (95 %) patients. EDV after automated segmentation: 119.1 ± 44.0ml; after ADJ-step 1: 115.8 ± 39.5 ml; after ADJ-step 2: 116.2 ± 39.4 ml. The corresponding results for ESV were 52.0 ± 18.5/49.6 ± 16.9/49.7 ± 16.4 ml; for SV 67.1 ± 28.5/66.2 ± 25.4/66.5 ± 25.5 ml; for EF 55.5 ± 7.3/56.7 ± 6.6/56.7 ± 6.3%; for MM 83.7 ± 35.9/76.2 ± 28.3/74.6 ± 27.2 g. Significant differences were found for ESV/MM/EF comparing the automated segmentation results with these after ADJ-step 1 and ADJ-step 2. No significant differences were found when comparing all results of ADJ-step 1 and ADJ-step 2 or when comparing EDV/SV results. Intra- and inter-rater reliability was excellent. The mean time effort was 63.4 ± 6.9 s for the automated segmentation, 74.2 ± 8.9 s for ADJ-step 1 and 269.5 ± 39.4 s for ADJ-step 2. Automated left ventricular volumes and function analysis in children and adolescents with surgically treated CHD proved to be feasible with excellent intra- and inter-rater reliability. Automated segmentation with manual apex/base adjustment provided

  8. OSH technical reference manual

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    In an evaluation of the Department of Energy (DOE) Occupational Safety and Health programs for government-owned contractor-operated (GOCO) activities, the Department of Labor`s Occupational Safety and Health Administration (OSHA) recommended a technical information exchange program. The intent was to share written safety and health programs, plans, training manuals, and materials within the entire DOE community. The OSH Technical Reference (OTR) helps support the secretary`s response to the OSHA finding by providing a one-stop resource and referral for technical information that relates to safe operations and practice. It also serves as a technical information exchange tool to reference DOE-wide materials pertinent to specific safety topics and, with some modification, as a training aid. The OTR bridges the gap between general safety documents and very specific requirements documents. It is tailored to the DOE community and incorporates DOE field experience.

  9. JANNAF 28th Propellant Development and Characterization Subcommittee and 17th Safety and Environmental Protection Subcommittee Joint Meeting. Volume 1

    Science.gov (United States)

    Cocchiaro, James E. (Editor); Mulder, Edwin J. (Editor); Gomez-Knight, Sylvia J. (Editor)

    1999-01-01

    This volume contains 37 unclassified/unlimited-distribution technical papers that were presented at the JANNAF 28th Propellant Development & Characterization Subcommittee (PDCS) and 17th Safety & Environmental Protection Subcommittee (S&EPS) Joint Meeting, held 26-30 April 1999 at the Town & Country Hotel and the Naval Submarine Base, San Diego, California. Volume II contains 29 unclassified/limited-distribution papers that were presented at the 28th PDCS and 17th S&EPS Joint Meeting. Volume III contains a classified paper that was presented at the 28th PDCS Meeting on 27 April 1999. Topics covered in PDCS sessions include: solid propellant rheology; solid propellant surveillance and aging; propellant process engineering; new solid propellant ingredients and formulation development; reduced toxicity liquid propellants; characterization of hypergolic propellants; and solid propellant chemical analysis methods. Topics covered in S&EPS sessions include: space launch range safety; liquid propellant hazards; vapor detection methods for toxic propellant vapors and other hazardous gases; toxicity of propellants, ingredients, and propellant combustion products; personal protective equipment for toxic liquid propellants; and demilitarization/treatment of energetic material wastes.

  10. Inventory of Federal energy-related environment and safety research for FY 1979. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    1980-12-01

    The FY 1979 Federal Inventory contains information on 3506 federally funded energy-related environmental and safety research projects. The Inventory is published in two volumes: Volume I, an executive summary and overview of the data and Volume II, project listings, summaries, and indexes. Research and development (R and D) categories were reorganized into three main areas; environmental and safety control technology, technology impacts overview and assessments, and biological and environmental R and D and assessments. Federal offices submitting project data were: Council on Environmental Quality; Department of Agriculture; Department of Commerce; Department of Defense; Department of Energy; Department of Health, Education, and Welfare; Department of Housing and Urban Development; Department of the Interior; Department of Transportation; Environmental Protection Agency; National Aeronautics and Space Administration; Nuclear Regulatory Commission; National Science Foundation; Office of Technology Assessment; and Tennessee Valley Authority. The inventory also breaks out research sponsored by various federal agencies and the amount of funding provided by each in various research categories. The format and index system allows efficient access to information compiled. Users are able to identify projects by log agency, performing organization, principal investigator and subject

  11. Nuclear Safety. Technical Progress Journal, October--December 1991: Volume 32, No. 4

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  12. Paul Scherrer Institute Scientific Report 2000. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Smith, Brian; Gschwend, Beatrice

    2001-03-01

    Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. Comprehensive assessments of energy systems are carried out in cooperation with PSI's General Energy Research Department. Many of the programs are part of collaborations with universities, industry, or international organisations. Progress in 2000 in these topical areas is described in this report. A list of scientific publications in 2000 is also provided

  13. Nuclear Safety. Technical Progress Journal, April--June 1993: Volume 34, No. 2

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1993-01-01

    This review journal that covers significant developments in the field of Nuclear Safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations are also treated.

  14. Nuclear Safety Technical Progress Journal, January--June 1995. Volume 36, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1995-01-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  15. Paul Scherrer Institute Scientific Report 2000. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Brian; Gschwend, Beatrice [eds.

    2001-03-01

    Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. Comprehensive assessments of energy systems are carried out in cooperation with PSI's General Energy Research Department. Many of the programs are part of collaborations with universities, industry, or international organisations. Progress in 2000 in these topical areas is described in this report. A list of scientific publications in 2000 is also provided.

  16. Nuclear Safety. Technical Progress Journal, July--September 1992: Volume 33, No. 3

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-01-01

    This review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  17. Flightfax: Army Aviation Risk-Management Information; Safety is a Leader's Job, Volume 27, Number 10

    National Research Council Canada - National Science Library

    1999-01-01

    .... However, what has changed in recent years has been our thinking about the human origins of accidents, Safety culture is one such concept that explicitly addresses the wider social causes of accidents...

  18. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    International Nuclear Information System (INIS)

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits

  19. NPP safety and personnel training. XII International conference. Abstracts. Volume 2

    International Nuclear Information System (INIS)

    2011-01-01

    The XII International conference NPP Safety and Personnel Training took place in Obninsk, October 4-7 2011. The problems of personnel training for nuclear industry are discussed. The innovation nuclear systems and fuel cycle are considered. The much attention has been given to NPP radiation safety and radioecology issues. The recent high-speed computation and simulation methods used in reactor technology are presented [ru

  20. Efficacy and safety of 10,600-nm carbon dioxide fractional laser on facial skin with previous volume injections

    Directory of Open Access Journals (Sweden)

    Josiane Hélou

    2013-01-01

    Full Text Available Background: Fractionated carbon dioxide (CO 2 lasers are a new treatment modality for skin resurfacing. The cosmetic rejuvenation market abounds with various injectable devices (poly-L-lactic acid, polymethyl-methacrylate, collagens, hyaluronic acids, silicone. The objective of this study is to examine the efficacy and safety of 10,600-nm CO 2 fractional laser on facial skin with previous volume injections. Materials and Methods: This is a retrospective study including 14 patients treated with fractional CO 2 laser and who have had previous facial volume restoration. The indication for the laser therapy, the age of the patients, previous facial volume restoration, and side effects were all recorded from their medical files. Objective assessments were made through clinical physician global assessment records and improvement scores records. Patients′ satisfaction rates were also recorded. Results: Review of medical records of the 14 patients show that five patients had polylactic acid injection prior to the laser session. Eight patients had hyaluronic acid injection prior to the laser session. Two patients had fat injection, two had silicone injection and one patient had facial thread lift. Side effects included pain during the laser treatment, post-treatment scaling, post-treatment erythema, hyperpigmentation which spontaneously resolved within a month. Concerning the previous facial volume restoration, no granulomatous reactions were noted, no facial shape deformation and no asymmetry were encountered whatever the facial volume product was. Conclusion: CO 2 fractional laser treatments do not seem to affect facial skin which had previous facial volume restoration with polylactic acid for more than 6 years, hyaluronic acid for more than 0.5 year, silicone for more than 6 years, or fat for more than 1.4 year. Prospective larger studies focusing on many other variables (skin phototype, injected device type are required to achieve better

  1. Inventory of Federal energy-related environment and safety research for FY 1978. Volume II. Project listings and indexes

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-12-01

    This volume contains summaries of FY-1978 government-sponsored environment and safety research related to energy. Project summaries were collected by Aerospace Corporation under contract with the Department of Energy, Office of Program Coordination, under the Assistant Secretary for Environment. Summaries are arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each agency. Information about the projects is included in the summary listings. This includes the project title, principal investigators, research organization, project number, contract number, supporting organization, funding level if known, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in Volume IV.

  2. Inventory of Federal energy-related environment and safety research for FY 1979. Volume II. Project listings and indexes

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-12-01

    This volume contains summaries of FY 1979 government-sponsored environment and safety research related to energy arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each contributing agency. Information elements included in the summary listings are project title, principal investigators, research organization, project number, contract number, supporting organization, funding level, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in the back of this volume.

  3. Inventory of Federal energy-related environment and safety research for FY 1978. Volume II. Project listings and indexes

    International Nuclear Information System (INIS)

    1979-12-01

    This volume contains summaries of FY-1978 government-sponsored environment and safety research related to energy. Project summaries were collected by Aerospace Corporation under contract with the Department of Energy, Office of Program Coordination, under the Assistant Secretary for Environment. Summaries are arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each agency. Information about the projects is included in the summary listings. This includes the project title, principal investigators, research organization, project number, contract number, supporting organization, funding level if known, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in Volume IV

  4. Inventory of Federal energy-related environment and safety research for FY 1979. Volume II. Project listings and indexes

    International Nuclear Information System (INIS)

    1980-12-01

    This volume contains summaries of FY 1979 government-sponsored environment and safety research related to energy arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each contributing agency. Information elements included in the summary listings are project title, principal investigators, research organization, project number, contract number, supporting organization, funding level, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in the back of this volume

  5. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Overview of the methodology. Vol. 1 of 9 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) including a CD-ROM comprising all volumes

    International Nuclear Information System (INIS)

    2008-11-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. This document follows the guidelines of the INPRO report 'Methodology for the assessment of innovative nuclear reactors and fuel cycles, Report of Phase 1B (first part) of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)', IAEA-TECDOC-1434 (2004), together with its previous report Guidance for the evaluation for innovative nuclear reactors and fuel cycles, Report of Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), IAEA-TECDOC-1362 (2003). This INPRO manual is comprised of an overview volume (laid out in this report), and eight additional volumes (available on a CD-ROM attached to the inside back cover of this report) covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). The overview volume sets out the philosophy of INPRO and a general discussion of the INPRO methodology. This overview volume discusses the relationship of INPRO with the UN concept of sustainability to demonstrate how the

  6. IAEA safeguards technical manual

    International Nuclear Information System (INIS)

    1982-03-01

    Part F of the Safeguards Technical Manual is being issued in three volumes. Volume 1 was published in 1977 and revised slightly in 1979. Volume 1 discusses basic probability concepts, statistical inference, models and measurement errors, estimation of measurement variances, and calibration. These topics of general interest in a number of application areas, are presented with examples drawn from nuclear materials safeguards. The final two chapters in Volume 1 deal with problem areas unique to safeguards: calculating the variance of MUF and of D respectively. Volume 2 continues where Volume 1 left off with a presentation of topics of specific interest to Agency safeguards. These topics include inspection planning from a design and effectiveness evaluation viewpoint, on-facility site inspection activities, variables data analysis as applied to inspection data, preparation of inspection reports with respect to statistical aspects of the inspection, and the distribution of inspection samples to more than one analytical laboratory. Volume 3 covers generally the same material as Volumes 1 and 2 but with much greater unity and cohesiveness. Further, the cook-book style of the previous two volumes has been replaced by one that makes use of equations and formulas as opposed to computational steps, and that also provides the bases for the statistical procedures discussed. Hopefully, this will help minimize the frequency of misapplications of the techniques

  7. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  8. Wien Automatic System Planning (WASP) Package. A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 1: Chapters 1-11

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). The present document has been produced to support use of the WASP-Ill Plus computer code and to illustrate the capabilities of the program. This Manual is organized in two separate volumes. This first one includes 11 main chapters describing how to use the WASP-Ill Plus computer program. Chapter 1 gives a summary description and some background information about the program. Chapter 2 introduces some concepts, mainly related to the computer requirements imposed by the program, that are used throughout the Manual. Chapters 3 to 9 describe how to execute each of the various programs (or modules) of the WASP-Ill Plus package. (abstract truncated)

  9. Wien Automatic System Planning (WASP) Package. A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 1: Chapters 1-11

    International Nuclear Information System (INIS)

    1995-01-01

    (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). The present document has been produced to support use of the WASP-Ill Plus computer code and to illustrate the capabilities of the program. This Manual is organized in two separate volumes. This first one includes 11 main chapters describing how to use the WASP-Ill Plus computer program. Chapter 1 gives a summary description and some background information about the program. Chapter 2 introduces some concepts, mainly related to the computer requirements imposed by the program, that are used throughout the Manual. Chapters 3 to 9 describe how to execute each of the various programs (or modules) of the WASP-Ill Plus package. The description for each module shows the user how to prepare the Job Control statements and input data needed to execute the module and how to interpret the printed output produced. The iterative process that should be followed in order to obtain the 'optimal solution' for a WASP case study is covered in Chapters 6 to 8. Chapter 10 explains the use of an auxiliary program of the WASP package which is mainly intended for saving computer time. Lastly, Chapter 11 recapitulates the use of WASP-Ill Plus for executing a generation expansion planning study; describes the several phases normally involved in this type of study; and provides the user with practical hints about the most important aspects that need to be verified at each phase while executing the various WASP modules

  10. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  11. Inventory of Federal energy-related environment and safety research for FY 1977. Volume IV. Indices

    Energy Technology Data Exchange (ETDEWEB)

    1978-07-01

    This volume contains indexes useful for accessing projects contained in the FY 1977 Federal Inventory. The indexing has been greatly broadened this year to provide hard copy users with greater flexibility in locating projects. The Inventory projects are printed sequentially by log number. An inventory log number is a unique number assigned to each project from a block of numbers set aside for each agency. The association of agencies with blocks of log numbers is found in the table of contents of the Index (Volume III).

  12. Aviation Support Equipment Technician (ASE 3 & 2 and ASM 3 & 2). Volume 1, Basics. Rate Training Manual and Nonresident Career Course.

    Science.gov (United States)

    Grant, Jesse R.; And Others

    This Rate Training Manual and Nonresident Career Course (RTM/NRCC) form a self-study package that will enable aviation support equipment personnel to help themselves fulfill the requirements for advancement. Designed for individual study and not formal classroom instruction, the RTM provides subject matter that relates directly to the occupational…

  13. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) data; Volume 5, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data.

  14. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  15. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual, Part 2: Human Error Probability (HEP) Data. Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  16. Treatment of emphysema using bronchoscopic lung volume reduction coil technology : an update on efficacy and safety

    NARCIS (Netherlands)

    Hartman, Jorine E.; Klooster, Karin; ten Hacken, Nick H. T.; Slebos, Dirk-Jan

    2015-01-01

    In the last decade several promising bronchoscopic lung volume reduction (BLVR) treatments were developed and investigated. One of these treatments is BLVR treatment with coils. The advantage of this specific treatment is that it works independently of collateral flow, and also shows promise for

  17. Communications Strategies on Alcohol and Highway Safety. Volume II. High School Youth. Final Report.

    Science.gov (United States)

    Grey Advertising, Inc., New York, NY.

    The second part of a two-part, two volume study deals with high school youth and identifies target populations and communications strategies for encouraging personal action steps to prevent drunk driving. Data, collected from interviews and questionnaires, are summarized and presented in tabular form. One fourth of high schoolers in a…

  18. Communications Strategies on Alcohol and Highway Safety. Volume I. Adults 18-55. Final Report.

    Science.gov (United States)

    Grey Advertising, Inc., New York, NY.

    The first part of a two-part, two volume study deals with adults aged 18-55 and identifies target populations and communications strategies for encouraging personal action steps to prevent drunk driving. Fully 54% of adult Americans participate once a month in social or business situations where alcohol is served. They are termed Alcohol Related…

  19. Communications strategies on alcohol and highway safety. Volume 1, Adults 18-55

    Science.gov (United States)

    1975-02-01

    The study is in two volumes, the first dealing with adults aged 18-55, the second with high school youth. Both identify target populations and communications strategies for encouraging personal action steps to prevent drunk driving. Fully 54% of adul...

  20. Technical background information for the environmental and safety report, Volume 4: White Oak Lake and Dam

    International Nuclear Information System (INIS)

    Oakes, T.W.; Kelly, B.A.; Ohnesorge, W.F.; Eldridge, J.S.; Bird, J.C.; Shank, K.E.; Tsakeres, F.S.

    1982-03-01

    This report has been prepared to provide background information on White Oak Lake for the Oak Ridge National Laboratory Environmental and Safety Report. The paper presents the history of White Oak Dam and Lake and describes the hydrological conditions of the White Oak Creek watershed. Past and present sediment and water data are included; pathway analyses are described in detail

  1. Paul Scherrer Institute Scientific Report 1999. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Smith, Brian; Gschwend, Beatrice

    2000-01-01

    Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The total effort invested in nuclear energy research in 1999 amounted to about 185 py/a and 4.7 MCHF of investment and maintenance costs. Approximately half of the salary, investment and maintenance costs are externally funded, primarily by the Swiss Utilities, the national co-operative for the disposal of nuclear waste (NAGRA), the Federal Office of Energy (BFE) through the nuclear safety inspectorate (HSK) and the Federal Office for Science and Education (BBW) in connection with the EU Framework Programmes; an increasing part of external funding is coming from domestic and foreign industry (nuclear component and fuel suppliers). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. 4 % of the total resources are invested in addressing more global aspects of energy. Many of the programs are part of collaborations with universities, industry, or international organisations. Progress in 1999 in these topical areas is described in this report. A list of scientific publications in 1999 is also provided

  2. Safety Research Experiment Facility Project. Conceptual design report. Volume II. Building and facilities

    International Nuclear Information System (INIS)

    1975-12-01

    The conceptual design of Safety Research Experiment Facility (SAREF) site system includes a review and evaluation of previous geotechnical reports for the area where SAREF will be constructed and the conceptual design of access and in-plant roads, parking, experiment-transport-vehicle maneuvering areas, security fencing, drainage, borrow area development and restoration, and landscaping

  3. Technical background information for the environmental and safety report, Volume 4: White Oak Lake and Dam

    Energy Technology Data Exchange (ETDEWEB)

    Oakes, T.W.; Kelly, B.A.; Ohnesorge, W.F.; Eldridge, J.S.; Bird, J.C.; Shank, K.E.; Tsakeres, F.S.

    1982-03-01

    This report has been prepared to provide background information on White Oak Lake for the Oak Ridge National Laboratory Environmental and Safety Report. The paper presents the history of White Oak Dam and Lake and describes the hydrological conditions of the White Oak Creek watershed. Past and present sediment and water data are included; pathway analyses are described in detail.

  4. A Handbook for Public Playground Safety. Volume II: Technical Guidelines for Equipment and Surfacing.

    Science.gov (United States)

    Consumer Product Safety Commission, Washington, DC.

    This handbook suggests safety guidelines for public playground equipment and describes various surfaces used under the equipment and possible injuries resulting from falls. The handbook is intended for use mainly by manufacturers, installers, school and park officials, and others interested in technical criteria for public playground equipment.…

  5. Nuclear safety: Volume 37, Number 3. Technical progress journal, July--September 1996

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, M D [ed.

    1996-09-01

    The seven articles in this journal are divided into the following areas: general safety considerations; design features; operating experiences; U.S. Nuclear Regulatory Commission information and analysis; and recent developments. Proposed rule changes since June 30, 1996 are also included.

  6. Drug research methodology. Volume 2, The identification of drugs of interest in highway safety

    Science.gov (United States)

    1980-03-01

    This report presents findings of a workshop on the identification of drugs that should be the focus of near-term highway safety research. Drugs of interest are those that have a potential to increase the likelihood of traffic crashes and their attend...

  7. Paul Scherrer Institute Scientific Report 1998. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Birchley, Jon; Ringele, Ruth

    1999-01-01

    Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The total effort invested in nuclear energy research in 1998 amounted to about 195 py/a and 4.5 millions CHF of investment and maintenance costs. Approximately half of the salary, investment and maintenance costs are externally funded, primarily by the Swiss Utilities, the national co-operative for the disposal of nuclear waste (NAGRA), the Federal Office of Energy (BFE) through the nuclear safety inspectorate (HSK) and the Federal Office for Science and Education (BBW) in connection with the EC Framework Programmes; an increasing part of external funding is coming from domestic and foreign industry (nuclear component and fuel suppliers). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. 4 % of the total resources are invested in addressing more global aspects of energy. Many of the programs are part of collaborations with universities, industry, or international organisations. A list of scientific publications in 1998 is included. (author)

  8. Paul Scherrer Institute Scientific Report 1998. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Birchley, Jon; Ringele, Ruth [eds.

    1999-09-01

    Nuclear energy related research in Switzerland is concentrated at PSI`s Nuclear Energy and Safety Research Department (NES). The total effort invested in nuclear energy research in 1998 amounted to about 195 py/a and 4.5 millions CHF of investment and maintenance costs. Approximately half of the salary, investment and maintenance costs are externally funded, primarily by the Swiss Utilities, the national co-operative for the disposal of nuclear waste (NAGRA), the Federal Office of Energy (BFE) through the nuclear safety inspectorate (HSK) and the Federal Office for Science and Education (BBW) in connection with the EC Framework Programmes; an increasing part of external funding is coming from domestic and foreign industry (nuclear component and fuel suppliers). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. 4 % of the total resources are invested in addressing more global aspects of energy. Many of the programs are part of collaborations with universities, industry, or international organisations. A list of scientific publications in 1998 is included. (author) figs., tabs., refs.

  9. Paul Scherrer Institute Scientific Report 1999. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Brian; Gschwend, Beatrice [eds.

    2000-07-01

    Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The total effort invested in nuclear energy research in 1999 amounted to about 185 py/a and 4.7 MCHF of investment and maintenance costs. Approximately half of the salary, investment and maintenance costs are externally funded, primarily by the Swiss Utilities, the national co-operative for the disposal of nuclear waste (NAGRA), the Federal Office of Energy (BFE) through the nuclear safety inspectorate (HSK) and the Federal Office for Science and Education (BBW) in connection with the EU Framework Programmes; an increasing part of external funding is coming from domestic and foreign industry (nuclear component and fuel suppliers). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. 4 % of the total resources are invested in addressing more global aspects of energy. Many of the programs are part of collaborations with universities, industry, or international organisations. Progress in 1999 in these topical areas is described in this report. A list of scientific publications in 1999 is also provided.

  10. GRACE manual

    International Nuclear Information System (INIS)

    Ishikawa, T.; Kawabata, S.; Shimizu, Y.; Kaneko, T.; Kato, K.; Tanaka, H.

    1993-02-01

    This manual is composed of three kinds of objects, theoretical background for calculating the cross section of elementary process, usage and technical details of the GRACE system. Throughout this manual we take the tree level process e + e - → W + W - γ as an example, including the e ± -scalar boson interactions. The real FORTRAN source code for this process is attached in the relevant sections as well as the results of calculation, which might be a great help for understanding the practical use of the system. (J.P.N.)

  11. Breckinridge Project, initial effort. Report VII, Volume 4. Safety and health plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    The Safety and Health Plan recognizes the potential hazards associated with the Project and has been developed specifically to respond to these risks in a positive manner. Prevention, the primary objective of the Plan, starts with building safety controls into the process design and continues through engineering, construction, start-up, and operation of the Project facilities and equipment. Compliance with applicable federal, state, and local health and safety laws, regulations, and codes throughout all Project phases is required and assured. The Plan requires that each major Project phase be thoroughly reviewed and analyzed to determine that those provisions required to assure the safety and health of all employees and the public, and to prevent property and equipment losses, have been provided. The Plan requires followup on those items or situations where corrective action needs were identified to assure that the action was taken and is effective. Emphasis is placed on loss prevention. Exhibit 1 provides a breakdown of Ashland Synthetic Fuels, Inc.'s (ASFI's) Loss Prevention Program. The Plan recognizes that the varied nature of the work is such as to require the services of skilled, trained, and responsible personnel who are aware of the hazards and know that the work can be done safely, if done correctly. Good operating practice is likewise safe operating practice. Training is provided to familiarize personnel with good operational practice, the general sequence of activities, reporting requirements, and above all, the concept that each step in the operating procedures must be successfully concluded before the following step can be safely initiated. The Plan provides for periodic review and evaluation of all safety and loss prevention activities at the plant and departmental levels.

  12. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  13. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  14. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  15. NRC safety research in support of regulation. Volume 8, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission`s rules, regulatory guides, or other guidance.

  16. Intra- and interoperator variability of lobar pulmonary volumes and emphysema scores in patients with chronic obstructive pulmonary disease and emphysema: comparison of manual and semi-automated segmentation techniques.

    Science.gov (United States)

    Molinari, Francesco; Pirronti, Tommaso; Sverzellati, Nicola; Diciotti, Stefano; Amato, Michele; Paolantonio, Guglielmo; Gentile, Luigia; Parapatt, George K; D'Argento, Francesco; Kuhnigk, Jan-Martin

    2013-01-01

    We aimed to compare the intra- and interoperator variability of lobar volumetry and emphysema scores obtained by semi-automated and manual segmentation techniques in lung emphysema patients. In two sessions held three months apart, two operators performed lobar volumetry of unenhanced chest computed tomography examinations of 47 consecutive patients with chronic obstructive pulmonary disease and lung emphysema. Both operators used the manual and semi-automated segmentation techniques. The intra- and interoperator variability of the volumes and emphysema scores obtained by semi-automated segmentation was compared with the variability obtained by manual segmentation of the five pulmonary lobes. The intra- and interoperator variability of the lobar volumes decreased when using semi-automated lobe segmentation (coefficients of repeatability for the first operator: right upper lobe, 147 vs. 96.3; right middle lobe, 137.7 vs. 73.4; right lower lobe, 89.2 vs. 42.4; left upper lobe, 262.2 vs. 54.8; and left lower lobe, 260.5 vs. 56.5; coefficients of repeatability for the second operator: right upper lobe, 61.4 vs. 48.1; right middle lobe, 56 vs. 46.4; right lower lobe, 26.9 vs. 16.7; left upper lobe, 61.4 vs. 27; and left lower lobe, 63.6 vs. 27.5; coefficients of reproducibility in the interoperator analysis: right upper lobe, 191.3 vs. 102.9; right middle lobe, 219.8 vs. 126.5; right lower lobe, 122.6 vs. 90.1; left upper lobe, 166.9 vs. 68.7; and left lower lobe, 168.7 vs. 71.6). The coefficients of repeatability and reproducibility of emphysema scores also decreased when using semi-automated segmentation and had ranges that varied depending on the target lobe and selected threshold of emphysema. Semi-automated segmentation reduces the intra- and interoperator variability of lobar volumetry and provides a more objective tool than manual technique for quantifying lung volumes and severity of emphysema.

  17. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P. [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines

  18. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines comprennent la seismicite

  19. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations

  20. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

  1. Fire Safety Aspects of Polymeric Materials. Volume 2. Test Methods, Specifications and Standards

    Science.gov (United States)

    1979-01-01

    Chairman: Dr. Seymour L. Blum Vice President Northern Energy Corporation 70 Memorial Drive Cambridge, MA 02142 Dr. George S. Ansell Dean, School...limitations, is perhaps the most firmly grounded and thoroughly docu- mented of any in the fire safety field (Benjamin and Adams , 1976). Having established...of the largest fire experiments ever undertaken. Operation Euroka, a 50 acre wildland fuel fire in Australia ( Adams et al., 1973). Scaling and

  2. Space shuttle/payload interface analysis. (Study 2.4) Volume 4: Business Risk and Value of Operations in Space (BRAVO). Part 2: User's manual

    Science.gov (United States)

    1974-01-01

    The BRAVO User's Manual is presented which describes the BRAVO methodology in terms of step-by-step procedures, so that it may be used as a tool for a team of analysts performing cost effectiveness analyses on potential future space applications. BRAVO requires a relatively general set of input information and a relatively small expenditure of resources. For Vol. 1, see N74-12493; for Vol. 2, see N74-14530.

  3. Shuttle user analysis (study 2.2). Volume 3: Business risk and value of operations in space (BRAVO). Part 2: User's manual

    Science.gov (United States)

    1974-01-01

    The purpose of the BRAVO User's Manual is to describe the BRAVO methodology in terms of step-by-step procedures. The BRAVO methodology then becomes a tool which a team of analysts can utilize to perform cost effectiveness analyses on potential future space applications with a relatively general set of input information and a relatively small expenditure of resources. An overview of the BRAVO procedure is given by describing the complete procedure in a general form.

  4. Waste Management Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, J.S. [ed.

    1967-08-31

    This Manual has been prepared to provide a documented compendium of the technical bases and general physical features of Isochem Incorporated`s Waste Management Program. The manual is intended to be used as a means of training and as a reference handbook for use by personnel responsible for executing the Waste Management Program. The material in this manual was assembled by members of Isochem`s Chemical Processing Division, Battelle Northwest Laboratory, and Hanford Engineering Services between September 1965 and March 1967. The manual is divided into the following parts: Introduction, contains a summary of the overall Waste Management Program. It is written to provide the reader with a synoptic view and as an aid in understanding the subsequent parts; Feed Material, contains detailed discussion of the type and sources of feed material used in the Waste Management Program, including a chapter on nuclear reactions and the formation of fission products; Waste Fractionization Plant Processing, contains detailed discussions of the processes used in the Waste Fractionization Plant with supporting data and documentation of the technology employed; Waste Fractionization Plant Product and Waste Effluent Handling, contains detailed discussions of the methods of handling the product and waste material generated by the Waste Fractionization Plant; Plant and Equipment, describes the layout of the Waste Management facilities, arrangement of equipment, and individual equipment pieces; Process Control, describes the instruments and analytical methods used for process control; and Safety describes process hazards and the methods used to safeguard against them.

  5. Site environmental report for 2000. Volume I, Environment, Health and Safety Division

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Robert [Environmental Services Group, Berkeley, CA (US); Javandel, Iraj [Environmental Services Group, Berkeley, CA (US); Lackner, Ginny [Environmental Services Group, Berkeley, CA (US); Ruggieri, Michael [Environmental Services Group, Berkeley, CA (US); Thorson, Patrick [Environmental Services Group, Berkeley, CA (US); Wahl, Linnea [Environmental Services Group, Berkeley, CA (US)

    2001-09-30

    Each year, Ernest Orlando Lawrence Berkeley National Laboratory (Berkeley Lab) prepared an integrated report on its environmental programs to satisfy the requirements of United States Department of Energy Order 231.1. The Site Environmental Report for 2000 is intended to summarize Berkeley Lab's compliance with environmental standards and requirements, characterize environmental management efforts through surveillance and monitoring activities, and highlight significant programs and efforts for calendar year 2000. Laboratory, the status of environmental programs, and summary results from surveillance and monitoring activities. Each chapter in Volume I begins with an outline of the sections that follow, including any tables or figures found in the chapter. Readers should use section numbers (e.g., §1.5) as navigational tools to find topics of interest in either the printed or the electronic version of the report. Volume II contains the individual data results from monitoring programs. Although a printed version of Volume II is not part of the report's initial distribution, it is available on request (see below). The report follows the Laboratory's policy of using the International System of Units (SI) or metric system of measurements. Whenever possible, results are also reported using the more conventional inch-pound system of measurements because this system is referenced by some current regulatory standards and may be more familiar to some readers. The tables included at the end of the Glossary are intended to help readers understand the various prefixes used with SI units of measurement and convert these units from one system to the other.

  6. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results.

  7. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  8. Atomic Safety and Licensing Board Panel Biennial Report, Fiscal Years 1993--1994. Volume 6

    International Nuclear Information System (INIS)

    1995-08-01

    In Fiscal Year 1993, the Atomic Safety and Licensing Board Panel (''the Panel'') handled 30 proceedings. In Fiscal Year 1994, the Panel handled 36 proceedings. The cases addressed issues in the construction, operation, and maintenance of commercial nuclear power reactors and other activities requiring a license form the Nuclear Regulatory Commission. This report sets out the Panel's caseload during the year and summarizes, highlight, and analyzes how the wide- ranging issues raised in those proceedings were addressed by the Panel's judges and licensing boards

  9. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

    International Nuclear Information System (INIS)

    1994-10-01

    The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results

  10. Atomic Safety and Licensing Board Panel Biennial Report, Fiscal Years 1993--1994. Volume 6

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    In Fiscal Year 1993, the Atomic Safety and Licensing Board Panel (``the Panel``) handled 30 proceedings. In Fiscal Year 1994, the Panel handled 36 proceedings. The cases addressed issues in the construction, operation, and maintenance of commercial nuclear power reactors and other activities requiring a license form the Nuclear Regulatory Commission. This report sets out the Panel`s caseload during the year and summarizes, highlight, and analyzes how the wide- ranging issues raised in those proceedings were addressed by the Panel`s judges and licensing boards.

  11. Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

    International Nuclear Information System (INIS)

    Ferrell, P.C.

    1997-01-01

    The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

  12. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

  13. Inventory of Federal Energy-Related Environment and Safety Research for FY 1978. Volume III, interactive terminal users guide

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C. E.; Barker, Janice F.

    1979-12-01

    This users' guide was prepared to provide interested persons access to, via computer terminals, federally funded energy-related environmental and safety research projects for FY 1978. Although this information is also available in hardbound volumes, this on-line searching capability is expected to reduce the time required to answer ad hoc questions and, at the same time, produce meaningful reports. The data contained in this data base are not exhaustive and represent research reported by the following agencies: Department of Agriculture, Department of Commerce, Department of Defense, Department of Energy, Department of Health, Education, and Welfare, Department of the Interior, Department of Transportation, Federal Energy Administration, National Aeronautics and Space Administration, National Science Foundation, Nuclear Regulatory Commission, Tennessee Valley Authority, U.S. Coast Guard, and the U.S. Environmental Protection Agency.

  14. Metrication manual

    International Nuclear Information System (INIS)

    Harper, A.F.A.; Digby, R.B.; Thong, S.P.; Lacey, F.

    1978-04-01

    In April 1978 a meeting of senior metrication officers convened by the Commonwealth Science Council of the Commonwealth Secretariat, was held in London. The participants were drawn from Australia, Bangladesh, Britain, Canada, Ghana, Guyana, India, Jamaica, Papua New Guinea, Solomon Islands and Trinidad and Tobago. Among other things, the meeting resolved to develop a set of guidelines to assist countries to change to SI and to compile such guidelines in the form of a working manual

  15. The environmental survey manual

    International Nuclear Information System (INIS)

    1987-08-01

    The purpose of this manual is to provide guidance to the Survey and Sampling and Analysis teams that conduct the one-time Environmental Survey of the major US Department of Energy (DOE) operating facilities. This manual includes a discussion of DOE's policy on environmental issues, a review of statutory guidance as it applies to the Survey, the procedures and protocols to be used by the Survey teams, criteria for the use of the Survey teams in evaluating existing environmental data for the Survey effort, generic technical checklists used in every Survey, health and safety guidelines for the personnel conducting the Survey, including the identification of potential hazards, prescribed protective equipment, and emergency procedures, the required formats for the Survey reports, guidance on identifying environmental problems that need immediate attention by the Operations Office responsible for the particular facility, and procedures and protocols for the conduct of sampling and analysis

  16. Evaluating the accuracy of the XVI dual registration tool compared with manual soft tissue matching to localise tumour volumes for post-prostatectomy patients receiving radiotherapy

    International Nuclear Information System (INIS)

    Campbell, Amelia; Brown, Elizabeth; Pryor, David; Lehman, Margot; Owen, Rebecca; Bernard, Anne

    2015-01-01

    Cone beam computerised tomography (CBCT) enables soft tissue visualisation to optimise matching in the post-prostatectomy setting, but is associated with inter-observer variability. This study assessed the accuracy and consistency of automated soft tissue localisation using XVI's dual registration tool (DRT). Sixty CBCT images from ten post-prostatectomy patients were matched using: (i) the DRT and (ii) manual soft tissue registration by six radiation therapists (RTs). Shifts in the three Cartesian planes were recorded. The accuracy of the match was determined by comparing shifts to matches performed by two genitourinary radiation oncologists (ROs). A Bland–Altman method was used to assess the 95% levels of agreement (LoA). A clinical threshold of 3 mm was used to define equivalence between methods of matching. The 95% LoA between DRT-ROs in the superior/inferior, left/right and anterior/posterior directions were −2.21 to +3.18 mm, −0.77 to +0.84 mm, and −1.52 to +4.12 mm, respectively. The 95% LoA between RTs-ROs in the superior/inferior, left/right and anterior/posterior directions were −1.89 to +1.86 mm, −0.71 to +0.62 mm and −2.8 to +3.43 mm, respectively. Five DRT CBCT matches (8.33%) were outside the 3-mm threshold, all in the setting of bladder underfilling or rectal gas. The mean time for manual matching was 82 versus 65 s for DRT. XVI's DRT is comparable with RTs manually matching soft tissue on CBCT. The DRT can minimise RT inter-observer variability; however, involuntary bladder and rectal filling can influence the tools accuracy, highlighting the need for RT evaluation of the DRT match.

  17. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-09-01

    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format

  18. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural & Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research & Regulatory Issues. Individual papers have been cataloged separately.

  19. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural ampersand Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research ampersand Regulatory Issues. Individual papers have been cataloged separately

  20. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately

  1. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume VI. Safety and environmental considerations for licensing

    International Nuclear Information System (INIS)

    1980-06-01

    This volume of the Nonproliferation Alternative Systems Assessment Program report addresses safety and environmental considerations in licensing the principal alternative nuclear reactors and fuel cycles in the United States for large-scale commercial nuclear power plants. In addition, this volume examines the safety and environmental considerations for licensing fuel service centers. These centers, which have been proposed for controlling sensitive fuel-cycle facilities and special nuclear materials, would contain a combination of such facilities as reprocessing plants, fabrication plants, and reactors. For this analysis, two fuel service center concepts were selected - one with power - generating capability and one without

  2. Selection methodology for LWR safety programs and proposals. Volume 2. Methodology application

    International Nuclear Information System (INIS)

    Ritzman, R.L.; Husseiny, A.A.

    1980-08-01

    The results of work done to update and apply a methodology for selecting (prioritizing) LWR safety technology R and D programs are described. The methodology is based on multiattribute utility (MAU) theory. Application of the methodology to rank-order a group of specific R and D programs included development of a complete set of attribute utility functions, specification of individual attribute scaling constants, and refinement and use of an interactive computer program (MAUP) to process decision-maker inputs and generate overall (multiattribute) program utility values. The output results from several decision-makers are examined for consistency and conclusions and recommendations regarding general use of the methodology are presented. 3 figures, 18 tables

  3. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  4. Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-01-01

    With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.

  5. NRC safety research in support of regulation - FY 1994. Volume 9

    International Nuclear Information System (INIS)

    1995-06-01

    This report, the tenth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1994. The goal of the Office of Nuclear Regulatory Research (RES) is to ensure the availability of sound technical bases for timely rulemaking and related decisions in support of NRC regulatory/licensing/inspection activities. RES also has responsibilities related to the resolution of generic safety issues and to the review of licensee submittals regarding individual plant examinations. It is the responsibility of RES to conduct the NRC's rulemaking process, including the issuance of regulatory guides and rules that govern NRC licensed activities

  6. Los Alamos National Laboratory corregated metal pipe saw facility preliminary safety analysis report. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-09-19

    This Preliminary Safety Analysis Report addresses site assessment, facility design and construction, and design operation of the processing systems in the Corrugated Metal Pipe Saw Facility with respect to normal and abnormal conditions. Potential hazards are identified, credible accidents relative to the operation of the facility and the process systems are analyzed, and the consequences of postulated accidents are presented. The risk associated with normal operations, abnormal operations, and natural phenomena are analyzed. The accident analysis presented shows that the impact of the facility will be acceptable for all foreseeable normal and abnormal conditions of operation. Specifically, under normal conditions the facility will have impacts within the limits posted by applicable DOE guidelines, and in accident conditions the facility will similarly meet or exceed the requirements of all applicable standards. 16 figs., 6 tabs.

  7. Paul Scherrer Institut Scientific Report 2001. Volume IV: Nuclear Energy and Safety

    International Nuclear Information System (INIS)

    Smith, B. L.

    2002-03-01

    The year 2001 was marked by the extensive audit of energy research at PSI. The audit took place on 21/22 June, and nuclear energy research was evaluated by five international experts. It was possible to present the quality and relevance of our research in a convincing way. This, together with optimal use of the available resources in our department Nuclear Energy and Safety (NES), prompted the highly welcome result that the auditors attested a high degree of scientific quality to the NES projects in general, with some of them judged to be of world-class standard. They recommended all audited projects be maintained, adequate resources to be allocated accordingly and, if necessary, an increase of public funding in nuclear energy research to be granted to achieve this. Finally, it was recommended that energy research should be explicitly mentioned in the statement of the ultimate mission of PSI. At the level of 'daily work', efforts in several directions related to the future of the Hotlab were one of the main focuses during the past year. On the one hand, the safety-related backfitting of the facility (particularly in regard to fire protection) has been continued, and was coupled with intensive planning studies, and with a thorough radiological cleaning of the labs; this invariably led to some impediment to the current research activities. Despite external burdens, which have led to some delays and additional costs, three of the four refurbishment phases have been completed, and the corresponding laboratories were brought back into operation after inspection and approval by the regulatory authority HSK. Given the size and complexity of the project, progress must be considered very satisfactory. A list of scientific publications in 2000 is also provided

  8. Safety report concerning Melusine (after power increase to 4 MW). Descriptive part. Volumes 1 and 2

    International Nuclear Information System (INIS)

    Baas, C.; Delcroix, V.; Jacquemain, M.; Marouby, R.; Meunier, C.; De Robien, E.; Rossillon, F.

    1967-03-01

    Construction of Melusine was started in January 1957, It first diverged on July 1, 1958. It operated at: 1 MW from March 1959; 1.4 MW from April 1960; 2 MW from September 1961; 4 MW from December 1965. Various modifications have been made since the reactor was built. They concern: - the addition of a hot cell. - the effluents: 2 reservoirs of 15 m 3 for liquid waste have been installed. The resin regeneration equipment has been completely modified. - the building: offices have been added - experimental zones have been set up in the hall - the electrical supply. - the cooling and purification circuits (installation of a second exchanger, replacement of the primary pumps, creation of a hot layer, etc... ). - the fuel elements (at the moment of the MTR type enriched to 90 per cent). - the swimming-pool (which has been partially equipped with a stainless steel coating). - the core (placing of 'stools', of a diving-board, etc...). - the ventilation: the hall has been de-pressurized during normal working in the event of an accident, the hall can be isolated and a safety circuit can be started up. A chimney has been installed. - the hall (which has been strengthened and sealed more effectively). - the control electronics (modification of the principle, and which are now entirely transistorized). So many changes have been made that the reactor now bears only a slight resemblance to the initial model. It has appeared necessary to make a brief review of these improvements in order to be able to judge more effectively the installations present safety characteristics; these latter are furthermore fairly well known as a result of the experiments carried out not only by the Thermal Transfer Service in Grenoble but also at Cadarache (Cabri) and Toulon (the work of Mr. PASCOUET). (authors) [fr

  9. Paul Scherrer Institut Scientific Report 2001. Volume IV: Nuclear Energy and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.L. (ed.)

    2002-03-01

    The year 2001 was marked by the extensive audit of energy research at PSI. The audit took place on 21/22 June, and nuclear energy research was evaluated by five international experts. It was possible to present the quality and relevance of our research in a convincing way. This, together with optimal use of the available resources in our department Nuclear Energy and Safety (NES), prompted the highly welcome result that the auditors attested a high degree of scientific quality to the NES projects in general, with some of them judged to be of world-class standard. They recommended all audited projects be maintained, adequate resources to be allocated accordingly and, if necessary, an increase of public funding in nuclear energy research to be granted to achieve this. Finally, it was recommended that energy research should be explicitly mentioned in the statement of the ultimate mission of PSI. At the level of 'daily work', efforts in several directions related to the future of the Hotlab were one of the main focuses during the past year. On the one hand, the safety-related backfitting of the facility (particularly in regard to fire protection) has been continued, and was coupled with intensive planning studies, and with a thorough radiological cleaning of the labs; this invariably led to some impediment to the current research activities. Despite external burdens, which have led to some delays and additional costs, three of the four refurbishment phases have been completed, and the corresponding laboratories were brought back into operation after inspection and approval by the regulatory authority HSK. Given the size and complexity of the project, progress must be considered very satisfactory. A list of scientific publications in 2000 is also provided.

  10. Safety Analysis in Large Volume Vacuum Systems Like Tokamak: Experiments and Numerical Simulation to Analyze Vacuum Ruptures Consequences

    Directory of Open Access Journals (Sweden)

    A. Malizia

    2014-01-01

    Full Text Available The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of hazardous internal material through the braches. It is clear that the influence of flow fields, consequence of accidents like LOVA, on dust resuspension is a key safety issue. In order to develop this analysis an experimental facility is been developed: STARDUST. This last facility has been used to improve the knowledge about LOVA to replicate a condition more similar to appropriate operative condition like to kamaks. By the experimental data the boundary conditions have been extrapolated to give the proper input for the 2D thermofluid-dynamics numerical simulations, developed by the commercial CFD numerical code. The benchmark of numerical simulation results with the experimental ones has been used to validate and tune the 2D thermofluid-dynamics numerical model that has been developed by the authors to replicate the LOVA conditions inside STARDUST. In present work, the facility, materials, numerical model, and relevant results will be presented.

  11. Accident Prevention: A Workers' Education Manual.

    Science.gov (United States)

    International Labour Office, Geneva (Switzerland).

    Devoted to providing industrial workers with a greater knowledge of precautionary measures undertaken and enforced by industries for the protection of workers, this safety education manual contains 14 lessons ranging from "The Problems of Accidents during Work" to "Trade Unions and Workers and Industrial Safety." Fire protection, safety equipment…

  12. Manual on indoor air quality

    International Nuclear Information System (INIS)

    Diamond, R.C.; Grimsrud, D.T.

    1983-12-01

    This reference manual was prepared to assist electric utilities in helping homeowners, builders, and new home buyers to understand a broad range of issues related to indoor air quality. The manual is directed to technically knowledgeable persons employed by utility companies - the customer service or marketing representative, applications engineer, or technician - who may not have specific expertise in indoor air quality issues. In addition to providing monitoring and control techniques, the manual summarizes the link between pollutant concentrations, air exchange, and energy conservation and describes the characteristics and health effects of selected pollutants. Where technical information is too lengthy or complex for inclusion in this volume, reference sources are given. Information for this manual was gathered from technical studies, manufacturers' information, and other materials from professional societies, institutes, and associations. The aim has been to provide objective technical and descriptive information that can be used by utility personnel to make informed decisions about indoor air quality issues

  13. Manual on indoor air quality

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, R.C.; Grimsrud, D.T.

    1983-12-01

    This reference manual was prepared to assist electric utilities in helping homeowners, builders, and new home buyers to understand a broad range of issues related to indoor air quality. The manual is directed to technically knowledgeable persons employed by utility companies - the customer service or marketing representative, applications engineer, or technician - who may not have specific expertise in indoor air quality issues. In addition to providing monitoring and control techniques, the manual summarizes the link between pollutant concentrations, air exchange, and energy conservation and describes the characteristics and health effects of selected pollutants. Where technical information is too lengthy or complex for inclusion in this volume, reference sources are given. Information for this manual was gathered from technical studies, manufacturers' information, and other materials from professional societies, institutes, and associations. The aim has been to provide objective technical and descriptive information that can be used by utility personnel to make informed decisions about indoor air quality issues.

  14. Safety study for HTR concepts under German site conditions. Main volume for phase I B

    International Nuclear Information System (INIS)

    1981-12-01

    Within the spectrum of accidents, core heat-up, caused by failure of the power supplies to the reactor or by interruption of the cooling system, is of major significance. This accident causes no serious effects to the environment if the reactor containment building remains intact; however, in certain infrequent events the containment building could fail due to overpressure after about one week at the earliest and a significant environmental hazard would ensue. The other accidents - depressurisation, water and air ingress - are by comparison of distinctly lower importance to the overall risk. A common feature of all the significant accidents is that within the first five hours the maximum release of radioactivity to the reactor containment building is the inventory of the primary circuit, while the fission product inventory of the core is securely retained. The accidents investigated are characterised in the study by the probability of occurrence, by the types and quantities of the fission products released and by the period of release. It was stipulated that in this phase of the study the health risk to the neighbouring population should not be determined. However, the estimated quantities of released nuclides and their associated time of release lead to the conclusion that the worst possible damage to the environment is very limited. This arises from the inherent safety characteristics of the HTR and the comparatively long period available for counter measures. (orig./HP) [de

  15. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

  16. Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

    International Nuclear Information System (INIS)

    1994-10-01

    This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G eff , is defined by: G eff - Σ M (F M x G M ) F M -fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials

  17. Optoelectronics circuits manual

    CERN Document Server

    Marston, R M

    1999-01-01

    This manual is a useful single-volume guide specifically aimed at the practical design engineer, technician, and experimenter, as well as the electronics student and amateur. It deals with the subject in an easy to read, down to earth, and non-mathematical yet comprehensive manner, explaining the basic principles and characteristics of the best known devices, and presenting the reader with many practical applications and over 200 circuits. Most of the ICs and other devices used are inexpensive and readily available types, with universally recognised type numbers.The second edition

  18. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

  19. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  20. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0: Models and Results Database (MAR-D) reference manual. Volume 8

    International Nuclear Information System (INIS)

    Russell, K.D.; Skinner, N.L.

    1994-07-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The primary function of MAR-D is to create a data repository for completed PRAs and Individual Plant Examinations (IPEs) by providing input, conversion, and output capabilities for data used by IRRAS, SARA, SETS, and FRANTIC software. As probabilistic risk assessments and individual plant examinations are submitted to the NRC for review, MAR-D can be used to convert the models and results from the study for use with IRRAS and SARA. Then, these data can be easily accessed by future studies and will be in a form that will enhance the analysis process. This reference manual provides an overview of the functions available within MAR-D and step-by-step operating instructions

  1. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  2. Program Management System manual

    International Nuclear Information System (INIS)

    1986-01-01

    The Program Management System (PMS), as detailed in this manual, consists of all the plans, policies, procedure, systems, and processes that, taken together, serve as a mechanism for managing the various subprograms and program elements in a cohesive, cost-effective manner. The PMS is consistent with the requirements of the Nuclear Waste Policy Act of 1982 and the ''Mission Plan for the Civilian Radioactive Waste Management Program'' (DOE/RW-0005). It is based on, but goes beyond, the Department of Energy (DOE) management policies and procedures applicable to all DOE programs by adapting these directives to the specific needs of the Civilian Radioactive Waste Management program. This PMS Manual describes the hierarchy of plans required to develop and maintain the cost, schedule, and technical baselines at the various organizational levels of the Civilian Radioactive Waste Management Program. It also establishes the management policies and procedures used in the implementation of the Program. These include requirements for internal reports, data, and other information; systems engineering management; regulatory compliance; safety; quality assurance; and institutional affairs. Although expanded versions of many of these plans, policies, and procedures are found in separate documents, they are an integral part of this manual. The PMS provides the basis for the effective management that is needed to ensure that the Civilian Radioactive Waste Management Program fulfills the mandate of the Nuclear Waste Policy Act of 1982. 5 figs., 2 tabs

  3. ETAP user's manual

    International Nuclear Information System (INIS)

    Watanabe, Norio; Higuchi, Suminori.

    1990-11-01

    The event tree analysis technique has been used in Probabilistic Safety Assessment for LWRs to delineate various accident scenarios leading to core melt or containment failure and to evaluate their frequencies. This technique often requires manual preparation of event trees with iterative process and time-consuming work in data handling. For the purpose of reducing manual efforts in event tree analysis, we developed a new software package named ETAP (Event Tree Analysis Supporting Program) for event tree analysis. ETAP is an interactive PC-based program which has the ability to construct, update, document, and quantify event trees. Because of its fast running capability to quantify event trees, use of the EATP program can make it easy to perform the sensitivity studies on a variety of system/containment performance issues. This report provides a user's manual for ETAP, which describes the structure, installation, and use of EATP. This software runs on NEC/PC-9800 or compatible PCs that have a 640 KB memory and MS-DOS 2.11 or higher. (author)

  4. School Survey on Crime and Safety (SSOCS): 2015-16. Public-Use Data File User's Manual. NCES 2018-107

    Science.gov (United States)

    Jackson, Michael; Diliberti, Melissa; Kemp, Jana; Hummel, Steven; Cox, Christina; Gbondo-Tugbawa, Komba; Simon, Dillon

    2018-01-01

    The School Survey on Crime and Safety (SSOCS) is managed by the National Center for Education Statistics (NCES) within the Institute of Education Sciences of the U.S. Department of Education. SSOCS collects extensive crime and safety data from principals and administrators of public schools in the United States. Data from this collection can be…

  5. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume VI. Safety and environmental considerations for licensing

    International Nuclear Information System (INIS)

    1979-12-01

    Volume 6 of the Nonproliferation Alternative Systems Assessment Program report addresses safety and environmental considerations in licensing the principal alternative nuclear reactors and fuel cycles in the United States for large-scale commercial nuclear power plants. In addition, this volume examines the safety and environmental considerations for licensing fuel service centers. These centers, which have been proposed for controlling sensitive fuel-cycle facilities and special nuclear materials, would contain a combination of such facilities as reprocessing plants, fabrication plants, and reactors. For this analysis, two fuel service center concepts were selected - one with power-generating capability and one without. This volume also provides estimates of the time required for development of large-scale commercial reactor systems to reach the construction permit application stage and for fuel-cycle facilities to reach the operating license application stage, which is a measure of the relative technical status of alternative nuclear systems

  6. Analytical methodology for safety validation of computer controlled subsystems. Volume 1 : state-of-the-art and assessment of safety verification/validation methodologies

    Science.gov (United States)

    1995-09-01

    This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety critical functions in high-speed rail or magnetic levitation ...

  7. Effects of MHRA drug safety advice on time trends in prescribing volume and indices of clinical toxicity for quinine

    Science.gov (United States)

    Acheampong, Paul; Cooper, Gill; Khazaeli, Behshad; Lupton, David J; White, Sue; May, Margaret T; Thomas, Simon H L

    2013-01-01

    Aims To ascertain the effects of the Medicines and Healthcare products Regulatory Agency's (MHRA) safety update in June 2010 on the volume of prescribing of quinine and on indices of quinine toxicity. Methods We analysed quarterly primary care total and quinine prescribing data for England and quinine prescribing volume for individual Primary Care Trusts in the North East of England from 2007/8 to 2011/12 obtained from the ePACT.net database. We also analysed quinine toxicity enquiries to the National Poisons Information Service (NPIS) via Toxbase® and by telephone between 2004/5 and 2011/12. Joinpoint regression and Pearson's correlation tests were used to ascertain changes in trends in prescribing and indices of toxicity and associations between prescribing and indices of toxicity, respectively. Results Total prescribing continued to increase, but annual growth in quinine prescribing in England declined from 6.0 to −0.6% following the MHRA update [difference −0.04 (95% confidence interval −0.07 to −0.01) quinine prescriptions per 100 patients per quarter, P = 0.0111]. Much larger reductions were observed in Primary Care Trusts that introduced comprehensive prescribing reviews. The previously increasing trend in Toxbase® quinine searches was reversed [difference −19.76 (95% confidence interval −39.28 to −9.20) user sessions per quarter, P = 0.0575]. Telephone enquiries to NPIS for quinine have declined, with stabilization of the proportion of moderate to severe cases of quinine poisoning since the update. Conclusions The MHRA advice was followed by limited reductions in the growth in quinine prescribing and in indicators of quinine overdose and toxicity. Quinine prescribing, however, remains common, and further efforts are needed to reduce availability and use. PMID:23594200

  8. COMMIX-1A: a three-dimensional transient single-phase computer program for thermal hydraulic analysis of single and multicomponent systems. Volume I: users manual

    International Nuclear Information System (INIS)

    Domanus, H.M.; Schmitt, R.C.; Sha, W.T.; Shah, V.L.

    1983-12-01

    The COMMIX-1A computer program is an updated and improved version of COMMIX-1 designed to analyze steady-state/transient, single-phase, three-dimensional fluid flow with heat transfer in reactor components and multicomponent systems. A new porous-media formulation via local volume averaging has been derived and employed in the COMMIX code. The concepts of volume porosity, directional surface permeability, distributed resistance, and distributed heat source or sink is used in the new porous-media formulation to model a flow domain with stationary structures. The concept of directional surface permeability is new and greatly facilitates modeling of velocity and temperature fields in anisotropic media. The new porous-media formulation represents the first unified approach to thermal-hydraulic analysis. It is now possible to perform a multidimensional thermal-hydraulic simulation of either a single component, such as a rod bundle, reactor plenum, piping system, heat exchanger, etc., or a multicomponent system that is a combination of these components. The conservation equations of mass, momentum, and energy based on the new porous-media formulation are solved as a boundary-value problem in space and an initial-value problem in time. Two other unique features provided in the COMMIX-1A code are (1) two solution procedures - a semi-implicit procedure modified from ICE and a fully-implicit procedure, named SIMPLEST-ANL, similar to the SIMPLE/SIMPLER algorithms - available a user's option and (2) a geometrical package capable of approximating many geometries. This report (Volume I) describes in detail the basic equations, formulations, solution procedures, flow charts, rebalancing scheme for faster convergence, options available to users, models to describe the auxiliary phenomena, input instructions, and two sample problems. The Volume II assembles and summarizes the results of many simulations that have been performed with COMMIX-1A computer program

  9. Pdap Manual

    DEFF Research Database (Denmark)

    Pedersen, Mads Mølgaard; Larsen, Torben J.

    Pdap, Python Data Analysis Program, is a program for post processing, analysis, visualization and presentation of data e.g. simulation results and measurements. It is intended but not limited to the domain of wind turbines. It combines an intuitive graphical user interface with python scripting...... that allows automation and implementation of custom functions. This manual gives a short introduction to the graphical user interface, describes the mathematical background for some of the functions, describes the scripting API and finally a few examples on how automate analysis via scripting is presented....... The newest version, and more documentation and help on how to used, extend and automate Pdap can be found at the webpage www.hawc2.dk...

  10. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

  11. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

  12. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  13. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

  14. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-06-01

    A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

  15. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  16. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  17. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    International Nuclear Information System (INIS)

    Hagrman, D.T.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

  18. Caltrans : construction manual

    Science.gov (United States)

    2009-08-01

    Caltrans intends this manual as a resource for all personnel engaged in contract administration. The manual establishes policies and procedures for the construction phase of Caltrans projects. However, this manual is not a contract document. It impos...

  19. Efficacy and safety of tolvaptan in heart failure patients with sustained volume overload despite the use of conventional diuretics: a phase III open-label study.

    Science.gov (United States)

    Fukunami, Masatake; Matsuzaki, Masunori; Hori, Masatsugu; Izumi, Tohru

    2011-12-01

    Volume overload is a common complication associated with heart failure (HF) and is recommended to be treated with loop or thiazide diuretics. However, use of diuretics can cause serum electrolyte imbalances and diuretic resistance. Tolvaptan, a selective, oral, non-peptide vasopressin V2-receptor antagonist, offers a new option for treating volume overload in HF patients. The aim of this study was to investigate the efficacy and safety of tolvaptan in Japanese HF patients with volume overload. Fifty-one HF patients with volume overload, despite using conventional diuretics, were treated with 15 mg/day tolvaptan for 7 days. If the response was insufficient at Day 7, tolvaptan was continued for a further 7 days at either 15 mg/day or 30 mg/day. Outcomes included changes in body weight, symptoms and safety parameters. Thirty-six patients discontinued treatment within 7 days, therefore 15 patients entered the second phase of treatment. In two patients, tolvaptan was increased to 30 mg/day after 7 days. Body weight was reduced on Day 7 (-1.95 ± 1.98 kg; n = 41) and Day 14 (-2.35 ± 1.44 kg; n = 11, 15 mg/day). Symptoms of volume overload, including lower limb edema, pulmonary congestion, jugular venous distention and hepatomegaly, were improved by tolvaptan treatment for 7 or 14 days. Neither tolvaptan increased the incidence of severe or serious adverse events when administered for 7-14 days. This study confirms the efficacy and safety of 15 mg/day tolvaptan for 7-14 days in Japanese HF patients with volume overload despite conventional diuretics.

  20. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  1. SCANS (Shipping Cask ANalysis System) a microcomputer-based analysis system for shipping cask design review: User's manual to Version 3a. Volume 1, Revision 2

    International Nuclear Information System (INIS)

    Mok, G.C.; Thomas, G.R.; Gerhard, M.A.; Trummer, D.J.; Johnson, G.L.

    1998-03-01

    SCANS (Shipping Cask ANalysis System) is a microcomputer-based system of computer programs and databases developed at the Lawrence Livermore National Laboratory (LLNL) for evaluating safety analysis reports on spent fuel shipping casks. SCANS is an easy-to-use system that calculates the global response to impact loads, pressure loads and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. SCANS is based on microcomputers compatible with the IBM-PC family of computers. The system is composed of a series of menus, input programs, cask analysis programs, and output display programs. All data is entered through fill-in-the-blank input screens that contain descriptive data requests. Analysis options are based on regulatory cases described in the Code of Federal Regulations 10 CFR 71 and Regulatory Guides published by the US Nuclear Regulatory Commission in 1977 and 1978

  2. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 3: Thermal hydraulic research and codes; Digital instrumentation and control; Structural performance

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-04-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) thermal hydraulic research and codes; (2) digital instrumentation and control; (3) structural performance

  3. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  4. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  5. HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, D.R.

    1999-10-07

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  6. WAM-E user's manual

    International Nuclear Information System (INIS)

    Rayes, L.G.; Riley, J.E.

    1986-07-01

    The WAM-E series of mainframe computer codes have been developed to efficiently analyze the large binary models (e.g., fault trees) used to represent the logic relationships within and between the systems of a nuclear power plant or other large, multisystem entity. These codes have found wide application in reliability and safety studies of nuclear power plant systems. There are now nine codes in the WAM-E series, with six (WAMBAM/WAMTAP, WAMCUT, WAMCUT-II, WAMFM, WAMMRG, and SPASM) classified as Type A Production codes and the other three (WAMFTP, WAMTOP, and WAMCONV) classified as Research codes. This document serves as a combined User's Guide, Programmer's Manual, and Theory Reference for the codes, with emphasis on the Production codes. To that end, the manual is divided into four parts: Part I, Introduction; Part II, Theory and Numerics; Part III, WAM-E User's Guide; and Part IV, WAMMRG Programmer's Manual

  7. Drug research methodology. Volume 1, The alcohol-highway safety experience and its applicability to other drugs

    Science.gov (United States)

    1980-03-01

    This report presents the findings of a workshop concerning the alcohol and highway safety experience, which includes research efforts to define the drinking-driving problem and societal responses to reduce the increased highway safety risk attributab...

  8. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    International Nuclear Information System (INIS)

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC's overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively

  9. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1

    International Nuclear Information System (INIS)

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC's overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively

  10. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  11. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  12. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  13. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 1: Plenary session; High burnup fuel; Containment and structural aging

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-01-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This first volume is divided into 3 sections: plenary session; high burnup fuel; and containment and structural aging. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  14. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  15. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  16. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  17. INIS: Database manual

    International Nuclear Information System (INIS)

    2003-01-01

    This document is one in a series of publications known as the INIS Reference Series. It is intended for users of INIS (International Nuclear Information System) output data on various media (FTP file, CD-ROM, e-mail file, earlier magnetic tape, cartridge, etc.). This manual provides a description of each data element including information on contents, structure and usage as well as historical overview of additions, deletions and changes of data elements and their contents that have taken place over the years. Each record contains certain control data fields (001-009), one, two or three bibliographic levels, a set of descriptors, and zero, one or more abstracts, one in English and optionally one or more in another language. In order to facilitate the description of the system, the sequence of data elements is based on the input or, as it is internally called, worksheet format which differs from the exchange format described in the manual IAEA-INIS-7. A separate section is devoted to each data element and deviations from the exchange format are indicated whenever present. As the Record Leader and the Directory are sufficiently explained in Chapter 3.1 of IAEA-INIS-7, the contents of this manual are limited to control fields and data fields; the detailed explanations are intended to supplement the basic information given in Chapter 3.2 of IAEA-INIS-7. Bibliographic levels are used to identify component parts of a publication, i.e. chapters in a book, articles in a journal issue, conference papers in a proceedings volume. All bibliographic levels contained in a record are given in a control data field. Each bibliographic level identifier appears in the subdirectory with a pointer to its position in the record

  18. ARDS User Manual

    Science.gov (United States)

    Fleming, David P.

    2001-01-01

    Personal computers (PCs) are now used extensively for engineering analysis. their capability exceeds that of mainframe computers of only a few years ago. Programs originally written for mainframes have been ported to PCs to make their use easier. One of these programs is ARDS (Analysis of Rotor Dynamic Systems) which was developed at Arizona State University (ASU) by Nelson et al. to quickly and accurately analyze rotor steady state and transient response using the method of component mode synthesis. The original ARDS program was ported to the PC in 1995. Several extensions were made at ASU to increase the capability of mainframe ARDS. These extensions have also been incorporated into the PC version of ARDS. Each mainframe extension had its own user manual generally covering only that extension. Thus to exploit the full capability of ARDS required a large set of user manuals. Moreover, necessary changes and enhancements for PC ARDS were undocumented. The present document is intended to remedy those problems by combining all pertinent information needed for the use of PC ARDS into one volume.

  19. Methane measurements manual; Handbok metanmaetningar

    Energy Technology Data Exchange (ETDEWEB)

    Holmgren, Magnus Andreas (SP Technical research institute of Sweden, Boraas (Sweden))

    2011-02-15

    Emissions to air in different parts of the system may arise in biogas plants, where there is biological treatment of organic matter by anaerobic degradation, and during upgrading of biogas to vehicle fuel. There are mainly four reasons why these emissions must be minimized. These are safety, greenhouse gas emissions, economy and smell. This manual gathers experience of several years of work with measurement of methane emissions from biogas and upgrading facilities. This work has been done mainly in the context of Swedish Waste Management's system of voluntary commitment. The purpose of this manual is to standardize methods and procedures when methane measurements are carried out so that the results are comparable between different providers. The main target group of the manual is measurement consultants performing such measurements. Calculation template in Excel is part of the manual, which further contributes to the measurements evaluated in a standardized way. The manual contains several examples which have been calculated in the accompanying Excel template. The handbook also contains a chapter mainly intended for facility staff, in which implementation of accurate leak detection is described, and where there are hints of a system of so-called intermediate inspections to detect leaks in time

  20. New Oxford style manual

    CERN Document Server

    2012-01-01

    The New Oxford Style Manual brings together two essential reference works in a single volume: New Hart's Rules and the New Oxford Dictionary for Writers and Editors. New Hart's Rules, Oxford's definitive guide to style, consists of 20 chapters that provide authoritative and expert advice on how to prepare copy for publication. Topics covered include how to use italic, roman, and other type treatments, numbers and dates, law and legal references, illustrations, notes and references, and bibliographies. The guidelines are complemented by the New Oxford Dictionary for Writers and Editors, which features 25,000 alphabetically arranged entries giving authoritative advice on those words and names which raise questions time and time again because of spelling, capitalization, hyphenation, or cultural and historical context. Entries give full coverage of recommended spellings, variant forms, confusable words, hyphenation, capitalization, foreign and specialist terms, proper names, and abbreviations. The dictionary a...

  1. The role of informal dimensions of safety in high-volume organisational routines: an ethnographic study of test results handling in UK general practice.

    Science.gov (United States)

    Grant, Suzanne; Checkland, Katherine; Bowie, Paul; Guthrie, Bruce

    2017-04-27

    The handling of laboratory, imaging and other test results in UK general practice is a high-volume organisational routine that is both complex and high risk. Previous research in this area has focused on errors and harm, but a complementary approach is to better understand how safety is achieved in everyday practice. This paper ethnographically examines the role of informal dimensions of test results handling routines in the achievement of safety in UK general practice and how these findings can best be developed for wider application by policymakers and practitioners. Non-participant observation was conducted of high-volume organisational routines across eight UK general practices with diverse organisational characteristics. Sixty-two semi-structured interviews were also conducted with the key practice staff alongside the analysis of relevant documents. While formal results handling routines were described similarly across the eight study practices, the everyday structure of how the routine should be enacted in practice was informally understood. Results handling safety took a range of local forms depending on how different aspects of safety were prioritised, with practices varying in terms of how they balanced thoroughness (i.e. ensuring the high-quality management of results by the most appropriate clinician) and efficiency (i.e. timely management of results) depending on a range of factors (e.g. practice history, team composition). Each approach adopted created its own potential risks, with demands for thoroughness reducing productivity and demands for efficiency reducing handling quality. Irrespective of the practice-level approach adopted, staff also regularly varied what they did for individual patients depending on the specific context (e.g. type of result, patient circumstances). General practices variably prioritised a legitimate range of results handling safety processes and outcomes, each with differing strengths and trade-offs. Future safety

  2. Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 3, Generic Safety Assessment of Repository in Crystalline Rocks

    International Nuclear Information System (INIS)

    Motiejunas, S.; Poskas, P.

    2005-01-01

    In this Volume a generic safety assessment of the repository for spent nuclear fuel in crystalline rock in Lithuania is presented. Modeling of safety relevant radionuclide release from the defected canister and their transport through the near field and far field was performed. Doses to humans due to released radionuclides in the well water were calculated and compared with the dose restrictions existing in Lithuania. For this stage of generic safety assessment only two scenarios were chosen: base scenario and canister defect scenario. KBS-3 concept developed by SKB for disposal of spent nuclear fuel in Sweden was chosen as prototype for repository in crystalline basement in Lithuania. The KBS-3H design with horizontal canister emplacement is proposed as a reference design for Lithuania

  3. Mobile luminance data collection system user manual.

    Science.gov (United States)

    2013-04-01

    While nighttime driving volumes are lower than daytime, the percentage of crashes with respect to volume is higher, so it is critical that transportation agencies do everything economically possible to improve nighttime driving safety. One crucial pa...

  4. Wien Automatic System Package (WASP). A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1995-01-01

    With several Member States, the IAEA has completed a new version of the WASP program, which has been called WASP-Ill Plus since it follows quite closely the methodology of the WASP-Ill model. The major enhancements in WASP-Ill Plus with respect to the WASP-Ill version are: increase in the number of thermal fuel types (from 5 to 10); verification of which configurations generated by CONGEN have already been simulated in previous iterations with MERSIM; direct calculation of combined Loading Order of FIXSYS and VARSYS plants; simulation of system operation includes consideration of physical constraints imposed on some fuel types (i.e., fuel availability for electricity generation); extended output of the resimulation of the optimal solution; generation of a file that can be used for graphical representation of the results of the resimulation of the optimal solution and cash flows of the investment costs; calculation of cash flows allows to include the capital costs of plants firmly committed or in construction (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). This second volume of the document to support use of the WASP-Ill Plus computer code consists of 5 appendices giving some additional information about the WASP-Ill Plus program. Appendix A is mainly addressed to the WASP-Ill Plus system analyst and supplies some information which could help in the implementation of the program on the user computer facilities. This appendix also includes some aspects about WASP-Ill Plus that could not be treated in detail in Chapters 1 to 11. Appendix B identifies all error and warning messages that may appear in the WASP printouts and advises the user how to overcome the problem. Appendix C presents the flow charts of the programs along with a brief description of the objectives and structure of each module. Appendix D describes the

  5. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  6. Minutes of the Explosives Safety Seminar (21st) Held at Houston, Texas on 28-30 August 1984. Volume 1

    Science.gov (United States)

    1984-08-01

    PHASE CRDCAE BGIN AE 01*LTESPD,3-1 AND DD ESTABLISHES FEASIBILITY FEASIBILITY 1391 SUBMITTED, NEE STUDY TO HHO CRDC CRDC SAFETY SITE SAFETY SAFETY PLAN... emission spectrum of the fuel. (Figure 6) -- i This slide indicates the emission spectrum of a typical hydrocarbon ------. ignition. The specific...radiate over the entire IR band. Typical are hot manifolds, boilers , processing vessels, engines and the sun itself. The background radiation from a heat

  7. Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume I

    International Nuclear Information System (INIS)

    2000-12-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

  8. Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume II

    International Nuclear Information System (INIS)

    2000-12-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

  9. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K J

    1979-01-01

    Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

  10. Guidance for implementing an environmental, safety and health assurance program. Volume 2. A model plan for environmental, safety and health staff audits and appraisals

    International Nuclear Information System (INIS)

    Ellingson, A.C.

    1980-09-01

    This is 1 of 15 documents designed to illustrate how an Environmental, Safety and Health (ES and H) Assurance Program may be implemented. The generic definition of ES and H Assurance Programs is given in a companion document entitled An Environmental, Safety and Health Assurance Program Standard. This document is concerned with ES and H audit and appraisal activities of an ES and H Staff Organization as they might be performed in an institution whose ES and H program is based upon the ES and H Assurance Program Standard. An annotated model plan for ES and H Staff audits and appraisals is presented and discussed

  11. Guidance for implementing an environmental, safety, and health assurance program. Volume 10. Model guidlines for line organization environmental, safety and health audits and appraisals

    International Nuclear Information System (INIS)

    Ellingson, A.C.

    1981-10-01

    This is 1 of 15 documents designed to illustrate how an Environmental, Safety and Health (ES and H) Assurance Program may be implemented. The generic definition of ES and H Assurance Programs is given in a companion document entitled An Environmental, Safety and Health Assurance Program Standard. The Standard specifies that the operational level of an institution must have an internal assurance function, and this document provides guidance for the audit/appraisal portion of the operational level's ES and H program. The appendixes include an ES and H audit checklist, a sample element rating guide, and a sample audit plan for working level line organization internal audits

  12. Guidance for implementing an environmental, safety, and health-assurance program. Volume 15. A model plan for line organization environmental, safety, and health-assurance programs

    Energy Technology Data Exchange (ETDEWEB)

    Ellingson, A.C.; Trauth, C.A. Jr.

    1982-01-01

    This is 1 of 15 documents designed to illustrate how an Environmental, Safety and Health (ES and H) Assurance Program may be implemented. The generic definition of ES and H Assurance Programs is given in a companion document entitled An Environmental, Safety and Health Assurance Program Standard. This particular document presents a model operational-level ES and H Assurance Program that may be used as a guide by an operational-level organization in developing its own plan. The model presented here reflects the guidance given in the total series of 15 documents.

  13. Mid-Face Volumization With Hyaluronic Acid: Injection Technique and Safety Aspects from a Controlled, Randomized, Double-Blind Clinical Study.

    Science.gov (United States)

    Prager, Welf; Agsten, Karla; Kravtsov, Maria; Kerscher, Prof Martina

    2017-04-01

    BACKGROUND: Injection of hyaluronic acid (HA) volumizing fillers in the malar area is intended for rejuvenation of the mid-face. The choice of products, depth, and technique of injection depends on the desired level of volume enhancement and practitioners' preferences. OBJECTIVE: To describe a volumizing injection technique in the scope of a controlled, randomized, double-blind, single-center, split-face clinical study. A total of 45 subjects with bilateral symmetrical moderate to severe volume loss in the malar area received a single 2 mL injection of CPM®-26 (Cohesive Polydensified Matrix®) on one side and VYC®-20 (VYCROSS®) on the contralateral side of the face. The same injection technique was applied for both sides of the face. Use of anesthetics, overcorrection, and touch-ups were not permitted. The investigator completed a product satisfaction questionnaire. Adverse events (AE) and injection-site reactions (ISRs) were reported during the study. RESULTS: The products were placed at the epiperiosteal depth in 88.9% (n=40), at the subdermal depth in 8.9% (n=4) and at both levels in 2.2% (n=1) of subjects. Fanning technique using cannulae was applied in most cases (97.8%, n=44). Results of the investigator satisfaction questionnaire allowed to characterize CPM-26 in comparison to other volumizing gels. Both study products were generally well tolerated. Local reactions were transient and of mild to moderate intensity, with the most frequent ones being redness, pain, and swelling. CONCLUSION: Adequate injection technique in volumizing treatments is essential to create a natural aesthetic rejuvenation while respecting the safety aspect of the procedures. A 22G blunt cannula used with CPM-26 was preferred due to an easier and a more homogeneous distribution of the product. The investigator also appreciated CPM-26 for its ease of injection, positioning, lifting, and volumizing capacity. J Drugs Dermatol. 2017;16(4):351-357..

  14. Sedation and physiologic response to manual restraint after intranasal administration of midazolam in Hispaniolan Amazon parrots (Amazona ventralis).

    Science.gov (United States)

    Mans, Christoph; Guzman, David Sanchez-Migallon; Lahner, Lesanna L; Paul-Murphy, Joanne; Sladky, Kurt K

    2012-09-01

    Administration of intranasal midazolam (2 mg/kg) was evaluated for sedation and effects on cloacal temperature, respiratory rate, and heart rate in manually restrained Hispaniolan Amazon parrots (Amazona ventralis). Adult parrots (n=9) were administered either midazolam (2 mg/kg) or an equal volume of saline solution intranasally before a 15-minute manual restraint in a complete crossover study. Respiratory rate and sedation scores were recorded before and during capture and during and after 15 minutes of manual restraint. Heart rate and cloacal temperature were recorded during manual restraint. After restraint, the parrots received intranasal flumazenil (0.05 mg/kg) or an equal volume of saline solution, and the recovery time was recorded. In those birds that received midazolam, sedation was observed within 3 minutes of administration, and vocalization, flight, and defense responses were significantly reduced during capture. During manual restraint, the mean rate of cloacal temperature increase was significantly slower and remained significantly lower in birds that received midazolam compared with controls. Mean respiratory rates were significantly lower for up to 12 minutes in parrots that received midazolam compared with those receiving saline solution. Flumazenil antagonized the effects of midazolam within 10 minutes. No overt clinical adverse effects to intranasal midazolam and flumazenil administration were observed. Further studies on the safety of intranasal midazolam and flumazenil in this species are warranted.

  15. Feasibility and safety of cavity-directed stereotactic radiosurgery for brain metastases at a high-volume medical center

    Directory of Open Access Journals (Sweden)

    Paul Rava, MD PhD

    2016-07-01

    Conclusions: Excellent local control is achievable with cavity-directed SRS in well-selected patients, particularly for lesions with diameter <3 cm and resection cavity volumes <14 mL. Long-term survival is possible for select patients.

  16. Integrated Safety Management System Phase 1 and 2 Verification for the Environmental Restoration Contractor Volumes 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    CARTER, R.P.

    2000-04-04

    DOE Policy 450.4 mandates that safety be integrated into all aspects of the management and operations of its facilities. The goal of an institutionalized Integrated Safety Management System (ISMS) is to have a single integrated system that includes Environment, Safety, and Health requirements in the work planning and execution processes to ensure the protection of the worker, public, environment, and the federal property over the life cycle of the Environmental Restoration (ER) Project. The purpose of this Environmental Restoration Contractor (ERC) ISMS Phase MI Verification was to determine whether ISMS programs and processes were institutionalized within the ER Project, whether these programs and processes were implemented, and whether the system had promoted the development of a safety conscious work culture.

  17. The Coast Guard Proceedings of the Marine Safety and Security Council. Volume 72, Number 4, Winter 2015-2016

    Science.gov (United States)

    2016-04-13

    National Drug Threat Assessment reported that the annual economic cost of controlled prescription drug non-medical use (use beyond that of...scenario: On a clear Tuesday eve- ning, an outbound small coastal freight ship collides with an inbound offshore supply vessel while navigating a bend...NMCProceedings@uscg.mil Website: www.uscg.mil/proceedings Safety Management Systems Waterways: America’s Economic Engine Recreational Boating Safety i

  18. Manual of radioisotope production

    International Nuclear Information System (INIS)

    1966-01-01

    The Manual of Radioisotope Production has been compiled primarily to help small reactor establishments which need a modest programme of radioisotope production for local requirements. It is not comprehensive, but gives guidance on essential preliminary considerations and problems that may be met in the early stages of production. References are included as an aid to the reader who wishes to seek further in the extensive literature on the subject. In preparing the Manual, which is in two parts, the Agency consulted several Member States which already have long experience in radioisotope production. An attempt has been made to condense this experience, firstly, by setting out the technical and economic considerations which govern the planning and execution of an isotope programme and, secondly, by providing experimental details of isotope production processes. Part I covers topics common to all radioisotope processing, namely, laboratory design, handling and dispensing of radioactive solutions, quality control, measurement and radiological safety. Part II contains information on the fifteen radioisotopes in most common use. These are bromine-82, cobalt-58, chromium-51, copper-64, fluorine-18, gold-198, iodine-131, iron-59, magnesium-28, potassium-42, sodium-24, phosphorus-32, sulphur-35, yttrium-90 and zinc-65. Their nuclear properties are described, references to typical applications are given and published methods of production are reviewed; also included are descriptions in detail of the production processes used at several national atomic energy organizations. No attempt has been made to distinguish the best values for nuclear data or to comment on the relative merits of production processes. Each process is presented essentially as it was described by the contributor on the understanding that critical comparisons are not necessary for processes which have been well tried in practical production for many years. The information is presented as a guide to enable

  19. TA-55 change control manual

    International Nuclear Information System (INIS)

    Blum, T.W.; Selvage, R.D.; Courtney, K.H.

    1997-11-01

    This manual is the guide for initiating change at the Plutonium Facility, which handles the processing of plutonium as well as research on plutonium metallurgy. It describes the change and work control processes employed at TA-55 to ensure that all proposed changes are properly identified, reviewed, approved, implemented, tested, and documented so that operations are maintained within the approved safety envelope. All Laboratory groups, their contractors, and subcontractors doing work at TA-55 follow requirements set forth herein. This manual applies to all new and modified processes and experiments inside the TA-55 Plutonium Facility; general plant project (GPP) and line item funded construction projects at TA-55; temporary and permanent changes that directly or indirectly affect structures, systems, or components (SSCs) as described in the safety analysis, including Facility Control System (FCS) software; and major modifications to procedures. This manual does not apply to maintenance performed on process equipment or facility SSCs or the replacement of SSCs or equipment with documented approved equivalents

  20. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 1 of 2: Technical standard

    International Nuclear Information System (INIS)

    1998-05-01

    This Department of Energy (DOE) technical standard (referred to as the Standard) provides guidance for integrating and enhancing worker, public, and environmental protection during facility disposition activities. It provides environment, safety, and health (ES and H) guidance to supplement the project management requirements and associated guidelines contained within DOE O 430.1A, Life-Cycle Asset Management (LCAM), and amplified within the corresponding implementation guides. In addition, the Standard is designed to support an Integrated Safety Management System (ISMS), consistent with the guiding principles and core functions contained in DOE P 450.4, Safety Management System Policy, and discussed in DOE G 450.4-1, Integrated Safety Management System Guide. The ISMS guiding principles represent the fundamental policies that guide the safe accomplishment of work and include: (1) line management responsibility for safety; (2) clear roles and responsibilities; (3) competence commensurate with responsibilities; (4) balanced priorities; (5) identification of safety standards and requirements; (6) hazard controls tailored to work being performed; and (7) operations authorization. This Standard specifically addresses the implementation of the above ISMS principles four through seven, as applied to facility disposition activities